ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 88

DOE Review Release 08/31/2012

Document Title:

Nevada Test Site – Occupational 

Environmental Dose

Document Number: ORAUT-TKBS-0008-4 

Revision: 03

Effective Date: 08/24/2012

Type of Document: TBD

Supersedes: Revision 02

Subject Expert(s): Eugene M. Rollins

Approval: Signature on File Approval Date: 08/21/2012

Eugene M. Rollins, Document Owner

Concurrence: Signature on File Concurrence Date: 08/21/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 08/21/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 08/23/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 08/24/2012

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 2 of 88

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

04/20/2004 00 New technical basis document for the Nevada Test Site–

Occupational Environmental Dose. First approved issue. Initiated by 

Eugene M. Rollins.

12/08/2006 00 PC-1 Approved page change revision as a result of biennial review. Adds 

acronyms and abbreviations on pages 4 and 5. Updates required 

language on page 6 in the Introduction. Adds Purpose and Scope 

sections on page 7. Added instructions to dose reconstructors on 

pages 44 and 45 in Section 4.5. As a result of formal internal review, 

deletes Section 4.5.2 from page 45. Completes references on page 

47. This revision results in no change to the assigned dose and no 

PER is required. Training required: As determined by the Task 

Manager. Initiated by Eugene M. Rollins. Approval:

Signature on File 12/05/2006

Eugene M. Rollins, Document Owner

Signature on File 10/24/2006

John M. Byrne, Task 3 Manager

Signature on File 10/24/2006

Edward F. Maher, Task 5 Manager

Signature on File 11/15/2006

Kate Kimpan, Project Director

Brant A. Ulsh Signature on File for 12/08/2006

James W. Neton, Associate Director for Science

05/27/2008 01 Approved revision initiated to apply only to post 1962 employment 

periods and all covered employees including those identified by job 

classification as a drillback operator prior to 1965. Also applies to 

covered employees involved with any of the 10 underground tests 

that resulted in unexpected release of radioactive materials. Revised 

Table 4.2.1.2.2.-2 (new Table 4-3) to reflect 2,000 hr/yr and 2400 

m3/yr breathing rate. Deleted Tables 4.2.1.2.2-3 and 4.2.1.2.2-4. 

Deleted Sections 4.2.1.2.3 and 4.2.1.2.4. Added Sections 4.2.1.2.3 

and 4.1.2.4 to introduce revised method for assigning environmental 

ambient inhalation intakes. Replaced Section 4.2.2 with ambient 

ingestion intake section. Revised Section 4.4 to include dose from 

radon and other alpha and beta emitting radionuclides and to include 

discussion of radon exposure in Gravel Gerties. Added Section 

4.4.3. Revised Section 4.4.4, Table 4.21 to increase radon WLM for 

G-Tunnel prior to 1984 and to maximize radon exposures for 

unidentified work locations. Added discussion of radon in Gravel 

Gerties. Added Section 4.4.5 to provide method for assigning 

inhalation and ingestion intakes for underground workers. Added 

Section 4.5.2 to provide instructions to dose reconstructors for 

assigning inhalation and ingestion ambient environmental intakes for 

all workers at the NTS. Occupational environmental dose will 

increase for all claims not previously worked using ORAUT-OTIB0018 or ORAUT-OTIB-0002. Also, cases of lung, ET1, and ET2 will 

need to be re-evaluated for increased radon exposures. 

Attachments A and B (formerly Rollins 2007, SRDB Ref ID: 37805) 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 3 of 88

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

were added to support the changes discussed in Section A.6.0 which 

included enriching the near-field environment with refractory 

elements. The 01-C version only added refractory elements back to 

the depleted refractory near-field environment given by the Hicks 

data (i.e., multiplied the Hicks refractory values by 2). The 01-D 

version enriched the near-field refractory environment by multiplying 

the Hicks refractory values by 4. Assumption of the enriched 

refractory near-field environment resulted in larger fission and 

activation product inhalation and ingestion correction factors (Tables 

4-8 and 4-12) and increased number of affected organs and the 

resultant doses (Tables 4-9 and 4-14). Incorporates formal internal 

and NIOSH review comments. Constitutes a total rewrite of the 

document. Training required: As determined by the Task Manager. 

Initiated by Eugene M. Rollins. 

04/15/2010 02 Revision initiated to incorporate definitions and directions for dose 

reconstruction for nonpresumptive cancers that are excluded from 

the 1963 through 1992 Special Exposure Cohort. Added Section 

4.1.2. Added years 2002 through 2008 ambient external doses to 

Tables 4-15 and 4-16. Incorporates formal internal and NIOSH 

review comments. Updated ORAUT references. Training required: 

As determined by the Objective Manager. Initiated by Eugene M. 

Rollins. 

08/24/2012 03 Revision extended measured ambient radiation doses in Tables 4-15 

and 4-16 to 2010 in Section 4.3.1. Also in Section 4.3.1, clarification 

was added to ensure that ambient dose was not to be assigned for 

years when no data was available; in particular, prior to 1958. 

Instructions clarified that for unmonitored workers likely to be 

exposed prior to 1958, the 50% coworker dose in Table 6-11 of 

ORAUT-TKBS-0008-6 (the NTS External TBD) should be assigned. 

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

Eugene M. Rollins. 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 4 of 88

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 8

4.1 Introduction ............................................................................................................................ 10

4.1.1 Purpose...................................................................................................................... 11

4.1.2 Special Exposure Cohort Petition (SEC) Information for the Nevada Test Site ........... 11

4.1.3 Scope......................................................................................................................... 12

4.2 Internal Inhalation and Ingestion Dose to Aboveground Workers ........................................... 12

4.2.1 Onsite Releases to Air................................................................................................ 12

4.2.1.1 Source Description....................................................................................... 12

4.2.1.1.1 Weapons Testing........................................................................ 12

4.2.1.1.2 Ground Seepage of Noble Gases ............................................... 14

4.2.1.1.3 Tunnel Operations ...................................................................... 14

4.2.1.1.4 Containment Ponds .................................................................... 14

4.2.1.1.5 Drillbacks.................................................................................... 14

4.2.1.1.6 Laboratories................................................................................ 14

4.2.1.1.7 Radioactive Waste Management Sites ....................................... 14

4.2.1.2 Atmospheric Radionuclide Concentrations................................................... 15

4.2.1.2.1 Annual Inhalation Intakes of Tritium............................................ 16

4.2.1.2.2 Annual Intakes of Plutonium ....................................................... 19

4.2.1.2.3 Annual Intakes of Other Radionuclides ....................................... 20

4.2.1.2.4 Scaling Factors for Inhalation Intakes ......................................... 20

4.2.1.2.5 Correction for Resuspension for Early Times After 

Atmospheric Tests ...................................................................... 20

4.2.1.2.6 Correction for Inhalation Dose from Short-Lived Fission 

and Activation Products .............................................................. 27

4.2.2 Annual Intakes from Ingestion .................................................................................... 28

4.3 Occupational External Dose................................................................................................... 31

4.3.1 Ambient Radiation ...................................................................................................... 31

4.3.2 Releases of Noble Gases ........................................................................................... 34

4.4 Internal Inhalation and Ingestion Dose to Underground Workers............................................ 36

4.4.1 Underground Activities................................................................................................ 36

4.4.2 Radon Measurements ................................................................................................ 37

4.4.3 Underground Radon Concentrations .......................................................................... 37

4.4.4 Underground Worker Exposure to Radon................................................................... 38

4.4.5 Underground Worker Exposure to Other Radionuclides ............................................. 41

4.5 Instruction to Dose Reconstructors for Assignment of Environmental Intakes ........................ 41

4.5.1 Assigning External Ambient Dose............................................................................... 41

4.5.2 Assigning Internal Ambient Dose................................................................................ 42

4.6 Attributions and Annotations .................................................................................................. 42

References ........................................................................................................................................ 43

Glossary ............................................................................................................................................ 51
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 5 of 88

ATTACHMENT A, AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA ..................... 58

ATTACHMENT B, TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION 

AND INGESTION INTAKES AND SOURCES OF OVERESTIMATED 

ORGAN DOSE..................................................................................................... 85
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 6 of 88

LIST OF TABLES

TABLE TITLE PAGE

4-1 HTO atmospheric concentrations by year and area with estimated maximum and 

average annual organ dose.................................................................................................... 18

4-2 Atmospheric concentrations of 239Pu for sampled areas......................................................... 21

4-3 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 22

4-4 Inventory of contaminated soil................................................................................................ 25

4-5 Radionuclide areal soil deposition decay, corrected to 1963 .................................................. 25

4-6 Abundance of radionuclides in NTS soils in relation to 239Pu decay, corrected to 1963 .......... 26

4-7 Scaled inhalation intakes corrected for early resuspension .................................................... 27

4-8 Organ-specific inhalation dose fission and activation product correction factors..................... 28

4-9 Inhalation dose from short-lived fission and activation products ............................................. 28

4-10 Radionuclide soil concentration by area................................................................................. 29

4-11 Area-specific and maximum annual ingestion rates................................................................ 29

4-12 Organ-specific ingestion fission and activation correction factors........................................... 30

4-13 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/year ........................... 30

4-14 Organ-specific doses from ingestion of fission and activation products .................................. 30

4-15 Average ambient background radiations levels by year.......................................................... 32

4-16 Ambient radiation by area ...................................................................................................... 32

4-17 85Kr average background and the Area 20 net concentrations and annual organ doses 

from 1983 through 1992......................................................................................................... 35

4-18 Results of experiment to determine the effects of ventilation conditions on RDCs in the 

G-Tunnel inclined drift............................................................................................................ 38

4-19 Radon daughter concentrations for N-Tunnel in 1991 and 1992............................................. 39

4-20 Radon daughter concentrations for P-Tunnel in 1991 and 1992............................................. 39

4-21 Annual 222Rn exposures and uncertainties for internal dose reconstruction for miners 

and tunnel workers.................................................................................................................40

A-1 Atmospheric concentrations of 239Pu for sampled areas......................................................... 61

A-2 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 62

A-3 Inventory of contaminated soil................................................................................................ 64

A-4 Radionuclide areal soil deposition.......................................................................................... 64

A-5 Radionuclide areal soil deposition decay, corrected to 1963 .................................................. 65

A-6 Abundance of radionuclides in NTS soils relative to 239Pu decay, corrected to 1963 .............. 66

A-7 Scaled inhalation intakes corrected for early resuspension in 1963........................................ 67

A-8 Far-field refractory enrichment ratios...................................................................................... 69

A-9 Organ-specific inhalation dose fission and activation product correction factors..................... 72

A-10 Inhalation dose from short-lived fission and activation products ............................................. 76

A-11 Radionuclide soil concentration by area................................................................................. 76

A-12 Area-specific and maximum annual ingestion rates................................................................ 77

A-13 Annual organ doses from ingestion of 238,238Pu and 241Am...................................................... 78

A-14 Organ-specific ingestion fission and activation correction factors........................................... 83

A-15 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/yr ............................... 83

A-16 Organ-specific doses from ingestion of fission and activation products .................................. 84

B-1 Total annual organ doses greater than 0.0001 rem from 30 years of inhalation and 

ingestion intakes .................................................................................................................... 86
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 7 of 88

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Air sampling stations in 1997 ................................................................................................. 17

4-2 Resuspension factor as a function of time after initial deposition............................................ 27

A-1 Resuspension factor as a function of time after initial deposition............................................ 67

A-2 SMALL BOY – 90Sr fraction of total dose lungs ...................................................................... 70

A-3 LITTLE FELLER I – 90Sr fraction of total dose to lungs........................................................... 71

A-4 TURK – 90Sr fraction of total dose to lungs............................................................................. 71

A-5 Strontium-90 fraction of total dose.......................................................................................... 72

A-6 Strontium-90 fraction of total dose.......................................................................................... 73

A-7 Strontium-90 fraction of total dose.......................................................................................... 73

A-8 Strontium-90 fraction of total dose.......................................................................................... 74

A-9 Strontium-90 fraction of total dose to the liver ........................................................................ 74

A-10 Strontium-90 fraction of total dose to the red marrow............................................................. 75

A-11 Strontium-90 fraction of total dose to bone surface ................................................................ 75

A-12 Strontium-90 fraction of total ingestion dose .......................................................................... 80

A-13 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 80

A-14 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 81

A-15 Strontium-90 fraction of total ingestion dose .......................................................................... 81

A-16 Strontium-90 fraction of ingestion dose to the upper large intestine ....................................... 82

A-17 Strontium-90 fraction of total ingestion dose .......................................................................... 82
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 8 of 88

ACRONYMS AND ABBREVIATIONS 

Bq becquerel

CFR Code of Federal Regulations

Ci curie

cm centimeter

d day

DOD U.S. Department of Defense

DOE U.S. Department of Energy

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EPA U.S. Environmental Protection Agency

ET Extra-thoracic region

ft foot

g gram

GSD geometric standard deviation

GZ Ground Zero

HPD Health Protection Department

hr hour

HT airborne tritium

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

km kilometer

L liter

lb pound

LLI lower large intestine

LLNL Lawrence Livermore National Laboratory

LN lymph node

LOS line of sight

m meter

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mi mile

min minute

mL milliliter

mR milliroentgen

mrem millirem

NIOSH National Institute for Occupational Safety and Health
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 9 of 88

NTS Nevada Test Site

pCi picocurie

PERM passive environmental radon monitor

POC probability of causation

RBM red bone marrow

RDC radon daughter concentration 

REECo Reynolds Electrical & Engineering Company

RPISU radon progeny integrating sampling units 

SEC Special Exposure Cohort

SI small intestine

SRDB Ref ID Site Research Database Reference Identification (number)

Sv sievert

TBD technical basis document

TH thoracic

TLD thermoluminescent dosimeter

TRU transuranic

TTR Tonopah Test Range

ULI upper large intestine

U.S.C. United States Code

WL working level

WLM working level month

yr year

μCi microcurie

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 10 of 88

4.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) restrict the “performance of duty”

referred to in 42 U.S.C. § 7384n(b) to nuclear weapons work (NIOSH 2010a).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2010a):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 11 of 88

4.1.1 Purpose

This technical basis document (TBD) discusses occupational environmental dose. Occupational 

environmental dose is the dose individuals received at the Nevada Test Site (NTS or Test Site) while 

outside operational facilities but on the site during work activities. These doses can be internal or 

external depending on the characteristics of the individual radionuclides. While inhalation of most 

radionuclides would cause a dose to various organs in the body, noble gases would primarily cause 

only an external dose because these inert radionuclides are not readily absorbed by the body. 

However, inhalation of radon will result in dose to the lungs and respiratory airways due to the 

subsequent deposition and decay of daughter products. Occupational environmental dose would be 

received by workers in the aboveground and underground working environments.

With the cessation of atmospheric testing at NTS in 1962, the greatest potential for environmental 

intakes of radioactive material in the aboveground environment results from the inhalation of 

radioactive particles that were resuspended from NTS soils into the atmosphere and from ingestion of 

soils that were previously contaminated by atmospheric nuclear weapons tests, reactor tests, and 

safety tests. The potential inhalation intakes can be estimated from air sampling data in the NTS 

annual environmental reports (see the Environmental Reports list in the References section) coupled 

with extensive soil contamination data gathered between 1983 and 1991 (McArthur and Kordas1983, 

1985; McArthur and Mead 1987, 1988, 1989; McArthur 1991). Because the air monitoring data were 

limited to gross alpha and beta measurements, tritium, and isotopes of plutonium (e.g., 238Pu, 239Pu, 

240Pu), inhalation intakes of other relatively long-lived radionuclides that have been identified in NTS 

soils (e.g., 241Am, 60Co, 137Cs, 90Sr, 152Eu, 154Eu, 155Eu) are scaled to those of plutonium based on their 

relative abundance in NTS soils. Ingestion intakes can be estimated by assuming consumption of 

contaminated NTS soils. To ensure that inhalation and ingestion intakes are not underestimated, the 

relative abundances of the long-lived radionuclides in NTS soils determined from the 1991 soil 

contamination data (McArthur 1991) were decay-corrected back to 1963. In addition, to ensure that 

intakes and resultant doses were not underestimated, correction factors were developed to account 

for potential exposures to short-lived fission and activation products based on test-specific data 

provided by Hicks (1981a,b,c,d). In addition, a correction factor was developed for inhalation intakes 

that accounts for the phenomenon of early resuspension (Anspaugh et al. 2002).

4.1.2 Special Exposure Cohort Petition (SEC) Information for the Nevada Test Site

The status of Special Exposure Cohort (SEC) petitions for NTS is:

Classes Added to the SEC

• Department of Energy (DOE) employees or DOE contractor or subcontractor employees who 

worked at the Nevada Test Site from January 27, 1951 through December 31, 1962 for a 

number of work days aggregating at least 250 work days, either solely under this employment 

or in combination with work days within the parameters (excluding aggregate work day 

requirements) established for other classes of employees included in the SEC, and who were 

monitored or should have been monitored (Leavitt 2006). 

• All employees of the Department of Energy, its predecessor agencies, and its contractors and 

subcontractors who worked at the Nevada Test Site from January 1, 1963 through December 

31, 1992 for a number of work days aggregating at least 250 work days, occurring either solely 

under this employment or in combination with work days within the parameters established for 

one or more other classes of employees included in the SEC (Sebelius 2010). 

NIOSH has determined, and the Secretary of Health and Human Services has concurred, that in the 

absence of bioassay results for the worker, internal doses cannot be reconstructed between 1951 and 

1962 inclusive for an Energy Employee (EE). Based on the SEC petition evaluation, internal dose is 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 12 of 88

not to be reconstructed for work before 1963 unless a worker has specific bioassay results that can be 

directly related to an event or incident. Any bioassay results in the DOE files for NTS workers before 

1963 should be assumed to be valid [HASL-300 (Harley 1976) procedures were used in the early 

bioassay program)]; therefore, these results can be used to evaluate internal dose. Much of the 

internal monitoring for individuals during the SEC period was event related. However, certain job 

classifications required routine monitoring. These included radiation safety personnel, industrial 

hygienist, and security personnel. 

NIOSH has determined, and the Secretary of Health and Human Services has concurred that 

NIOSH lacks sufficient information that would allow it to adequately estimate internal exposures 

during the period 1963 through 1992 (Sebelius 2010). NIOSH believes that the cessation of nuclear 

testing, coupled with the implementation of the 1993 NTS internal technical basis document that 

demonstrates NTS compliance with 10 CFR Part 835, supports NIOSH’s ability to bound internal dose 

for the evaluated class starting in 1993. 

Dose reconstruction guidance in this technical basis document (TBD) for periods before January 1, 

1993 is presented to provide a technical basis for partial dose reconstructions for nonpresumptive 

cancers not covered in the SEC classes through December 31, 1992. Although NIOSH found that it is 

not possible to bound total internal dose for the proposed classes, it intends to use internal and 

external monitoring data that might become available for an individual claim (and that can be 

interpreted using its existing dose reconstruction processes or procedures). Therefore, dose 

reconstructions for individuals employed at NTS during the period from 1951 through December 31, 

1992, but who do not qualify for inclusion in the SEC, can be performed using these data as 

appropriate. 

4.1.3 Scope

Section 4.2 discusses internal doses to aboveground workers from onsite releases to the air and 

resuspension of radioactive materials in soil, as well as from ingestion of contaminated soils. 

Section 4.3 describes external doses to workers from ambient radiation and releases of radioactive 

noble gases to air. Section 4.4 discusses internal dose to underground workers from inhalation of 

ambient concentrations of radioactive materials in the air, ingestion of contaminated materials, and 

exposure to radon. Section 4.5 provides instructions to dose reconstructors for assignment of 

environmental intakes. Section 4.6 presents attributions and annotations, which are indicated by 

bracketed callouts and used to identify the source, justification, or clarification of the associated 

information.

4.2 INTERNAL INHALATION AND INGESTION DOSE TO ABOVEGROUND WORKERS 

Section 4.2.1 discusses the internal dose for workers outside the facilities, as determined from air 

concentrations resulting from ground-level releases. Section 4.2.2 discusses the internal dose from 

resuspension of radioactive materials in the soil. 

4.2.1 Onsite Releases to Air

4.2.1.1 Source Description

4.2.1.1.1 Weapons Testing

NTS has been the primary location for the testing of nuclear explosives in the continental United 

States since 1951. Test programs have included atmospheric testing in the 1950s and early 1960s, 

earth-cratering experiments, and open-air nuclear reactor and rocket engine testing. Since the mid-
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 13 of 88

1960s, testing of nuclear devices has occurred underground in drilled vertical holes or in mined 

tunnels. No nuclear tests have been conducted since September 1992 (Black 1995). 

In all, more than 900 nuclear tests have taken place at the Test Site as part of these programs. One 

result of these tests is that the surface soils in many parts of NTS contain measurable amounts of 

several long-lived radionuclides. Almost all of the more than 100 aboveground tests contaminated the 

soil near ground zero. In addition, several underground tests were cratering experiments that threw 

radioactive rock and soil hundreds of feet, and some deeper underground tests vented radioactive 

material to the surface. A few safety tests, in which a nuclear device was destroyed by conventional 

explosives, scattered plutonium (and in some cases uranium) over the nearby ground. Further, there 

was fallout of radioactive debris from many tests over the northern and eastern parts of the Test Site 

(McArthur 1991).

Radiation levels at NTS have been monitored regularly, and safety officials have identified and fenced 

off areas where the soil is heavily contaminated. In many other areas, radionuclide levels are not high 

enough to warrant closing the area, but they are still above background. 

Atmospheric weapon and safety tests from 1951 and 1963 resulted in the release of about 

1.2 × 1010 Ci to the atmosphere (DOE 1996). Much of this activity was from relatively short-lived 

radionuclides that decayed in a matter of days or weeks. The volatile radionuclides (such as 

radioiodines, noble gases, and tritium) were diluted in the atmosphere and transported off site. 

However, much of the nonvolatile, long-lived radionuclides settled in the soils at various locations (see 

Table 4-10 in Section 4.2.2). These contaminated soils continue to represent a potential inhalation 

pathway to workers from resuspension of soils by wind and such mechanical activities as cleanup and 

remediation (McArthur 1991).

In 1963, nuclear weapons testing was moved underground to prevent the release of radionuclides to 

the atmosphere and achieve containment. Releases of radioactive material after an underground test 

are generally categorized with terms that describe the volume of material released and the conditions 

of the release (DOE 1996):

• Containment Failures: Containment failures are unintentional releases of radioactive matter to 

the atmosphere due to failure of the containment system. A prompt massive release, or one 

that occurs soon after a test, is a venting.

• Late-Time Seeps: Late-time seeps are small releases that occur days or weeks after a test 

when gases diffuse through pore spaces in the overlying rock and are drawn to the surface by 

decreases in atmospheric pressure.

• Controlled Tunnel Purging: A controlled tunnel purge is an intentional release to enable either 

recovery of experimental data and equipment or reuse of part of the tunnel system.

• Operational Release: Operational releases are small consequential releases that occur when 

core or gas samples are collected, or when a drillback hole is sealed.

At present, processing of radioactive materials at NTS includes only laboratory analyses. Handling of 

these materials includes only the transport and storage of nuclear explosive devices and the operation 

of a radioactive waste management site for low-level radioactive and mixed wastes. Monitoring and 

evaluation of the activities indicate that the potential sources of onsite radiation exposure are releases 

from the following sources (Black 1995): 

• Tritiated water vapor (HTO) from drainage containment ponds for E Tunnel in Area 12

• Onsite radioanalytical laboratories
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 14 of 88

• Area 3 and 5 waste facilities

• Other diffuse sources

The following sections describe effluent sources at NTS from information in Black (1995).

4.2.1.1.2 Ground Seepage of Noble Gases

Ground seepage can be enhanced when changes in ambient pressure pump small amounts of noble 

gases up through the overburden and into the atmosphere from the cavity created by a nuclear test. 

This process, sometimes referred to as atmospheric pumping, creates a diffuse source of radiological 

effluents. These area sources are rare and therefore not routinely monitored. The phenomenon is 

usually restricted to tests in the Pahute Mesa region of NTS. These seepages are from nuclear tests 

before 1993. 

4.2.1.1.3 Tunnel Operations

Nuclear tests occurred in mined tunnel complexes in the Rainier Mesa region. Because some tunnels 

were sealed in the mid-1990s, small amounts of contaminated water continue to drain from only one 

tunnel (Section 4.2.1.1.4). 

4.2.1.1.4 Containment Ponds

Water contaminated with radionuclides seeps from the tunnels in Area 12 and collects in containment 

ponds where some evaporates and some seeps into the soil. The only radiological contaminant that 

produces a measurable air emission from evaporation of the water is tritium (3

H) in the form of HTO. 

4.2.1.1.5 Drillbacks

After underground nuclear tests, core samples have been taken from the cavity that was formed by 

the detonation for analysis and diagnosis. This core sampling is accomplished by drilling into the area 

of interest and recovering samples using special equipment. Radioactive material can escape to the 

atmosphere during these operations. 

4.2.1.1.6 Laboratories

Reynolds Electrical & Engineering Company (REECo) conducted radiological analyses in Building 

650, and Los Alamos National Laboratory conducted similar analyses in Building 701 at Mercury, 

Nevada. Because these facilities have processed primarily environmental samples, very little 

radioactivity has passed through them. However, there was potential for radionuclides to be 

discharged to the atmosphere through the hood ventilation system during sample processing, 

particularly of spiked samples or from loss of radioactive standards that contained heavy water, 

radioiodines, or noble gases. 

4.2.1.1.7 Radioactive Waste Management Sites

Areas 3 and 5 contain sites for the disposal of low-level radioactive waste, and Area 5 contains sites 

for storage of transuranic and mixed transuranic wastes, as well as the Greater Confinement Disposal 

Test Unit and 12 accompanying boreholes (only a few contain waste). Disposal occurs in pits and 

trenches; concrete pads provide temporary storage of certain wastes. Area 5 is for packaged waste 

disposal only. The Waste Examination Facility houses a glovebox with high-efficiency particulate air 

filtration that is used to examine and repack transuranic (TRU) waste drums. No contamination has 

been released from glovebox operations to the environment. The drums, which have been sent to 

NTS from Lawrence Livermore National Laboratory (LLNL) in past years, are stored in the TRU Pad 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 15 of 88

Cover Building. Repacked drums will be sent to the Waste Isolation Pilot Plant. The facility is a 

diffuse source of radiological effluents. The only radioactive effluent that has been detected by the 

various samplers around the site is HTO in atmospheric moisture. The Area 3 low-level waste site is 

in a location where surface soil has been contaminated by deposited plutonium, and resuspension of 

this soil by wind and vehicular activity has resulted in detection of above-background levels of 

plutonium in nearby air samples.

4.2.1.2 Atmospheric Radionuclide Concentrations

In 1964, REECo established an environmental surveillance program at NTS to measure radiological 

conditions throughout the site without regard to nuclear tests (Glora and Brown 1964a). That is, the 

collected data were not to relate to specific tests but to general conditions of radiation. The short-term 

objective of the program was to minimize casual personnel exposure to radiation by locating and 

identifying localized radiological environmental conditions by type and quantity of contamination. The 

long-range objective of the program was to establish baseline environmental data that could provide a 

reference for comparison with subsequent test activities and radiation measurements (Glora and 

Brown 1964a).

The initial surveillance program included, over time, 12 permanent air-sampling stations in the most 

populated areas at NTS. The air samplers were low-volume Filter Queen samplers with 8- by 10-in. 

(Gelman Type E) glass-fiber filter papers. Operating times were determined by integrated electric 

timers with flow rates from calibrated rotometers. Typical flow rates varied from 3 to 6 ft3

/min; 

samples were collected weekly (Glora and Brown 1964b).

After the first reporting period (June 1964), positive-displacement Gast pumps, which were equipped 

with in-line total volume gas meters, replaced the Filter Queen samplers. In May and June 1965, the 

8- by 10-in. Gelman filters were replaced with a new sampler that used 4-in. diameter Whatman 41 

filter paper (Lewis, Glora, and Aoki 1965). The sampling rate of these samplers was about 3 ft3

/min. 

Therefore, the total volume of sampled air in a 7-day period was about 1 × 103

 m3. During this period, 

the number of sampling stations increased to 13 and caustic scrubbers were added for the detection 

of radioiodines.

Early particulate samples were typically analyzed only for gross alpha and gross beta. However, if 

gross beta concentrations exceeded 1 × 10-5 μCi/m3

, researchers conducted an analysis for 227Ac (the 

most hazardous beta emitter present). Because no historical evidence exists that 227Ac has been 

detected in air or soil samples, the assumption that unidentified beta emitters were 227Ac would be 

unreasonable and inappropriate. Therefore, for dose reconstruction, 90Sr is assumed to be the 

unidentified beta-emitting radionuclide with the highest internal dose impact (see Section 4.2.1.2.4 for 

details).

Environmental samples were analyzed for gross alpha and beta radioactivity by gas proportional 

counting. Air samples were analyzed with a Nuclear Chicago ULTRASCALER system with a 

9- by 10-in. detection chamber to accommodate the 8- by 10-in. filters. All other samples were

analyzed with a Beckman WIDEBETA system equipped with an automatic sample changer. This 

counting equipment and protocol allowed lower detection limits of about 2.9 × 10-9 and 

1.6 × 10-8 μCi/m3 for gross alpha and gross beta, respectively (Lewis, Glora, and Aoki 1965). By 

1978, the lower detection limit for gross beta counting had been reduced to 1 × 10-10 μCi/m3 (Lewis, 

Glora, and Aoki 1965).

Gross gamma screening was typically performed on samples that were prepared for gross beta 

counting. The screening was performed using a 5- by 5-in. NaI(Tl) detector that was coupled to a 

single-channel analyzer. Any samples that showed elevated gamma readings were sent to a 

multichannel analyzer for radionuclide identification (Lewis, Glora, and Aoki 1965). In the early 1970s, 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 16 of 88

gamma spectroscopy was performed using germanium-lithium [Ge(Li)] detectors and 2,000-channel 

analyzers with a lower detection limit of 5 × 10-9 μCi/m3

 (Lantz 1978a). After cessation of atmospheric 

testing at NTS in 1963, fission products were frequently measured in the atmosphere but were 

typically correlated with foreign atmospheric weapons testing (see Section 4.2.1.2.4).

In 1971, weekly air samples from a given station were batched on a monthly basis and subjected to 

radiochemical analysis for 239Pu. The procedure included acid dissolution with ion exchange recovery 

and electroplating onto stainless-steel discs, which was followed by alpha spectroscopy using solidstate surface barrier detectors (Lantz 1978a). This analysis provided a nominal minimum detection 

limit of 3 × 10-7 μCi/m3 (Lantz 1978a). Routine analysis for 238Pu started in 1989 (Wruble and 

McDowell 1990a).

In 1977, a separate sampler was designed for the collection of airborne tritium (HT) and HTO. The 

sampler was portable and capable of unattended operation for as long as 2 weeks in desert areas. A 

small electronic pump drew air into the sampler at about 0.5 L/min, and the HTO was removed from 

the air stream by a silica gel drying column. The dry air then passed through a catalytic converter with 

platinum to generate HTO from HT. Another drying column collected the vapor, in which a small 

volume of distilled water served as a trap for the HTO. Appropriate aliquots of condensed moisture 

were obtained by heating the silica gel. Counting via liquid scintillation techniques enabled the 

determination of HT and HTO activities. The typical minimum detection limit for this analysis was 

3 × 10-7 μCi/m3

 (Lantz 1978a). 

The number of air sampling stations increased over the years to a peak of 52 in 1989 (Wruble and 

McDowell 1990a). This number remained fairly constant until a gradual reduction began in 1995 

(Black and Townsend 1996). This reduction occurred primarily because of a gradual strategy shift 

from environmental monitoring to demonstration of compliance with National Emission Standards for 

Hazardous Air Pollutants as approved by the U.S. Environmental Protection Agency (EPA; i.e., 

40 CFR Part 61) (Black and Townsend 1997). Figure 4-1 shows the locations of the 48 sampling 

stations in 1997 (Black and Townsend 1998).

Since the late 1980s, the environmental surveillance program has routinely monitored atmospheric 

concentrations of tritium, 238Pu, 239Pu, and 240Pu. These radionuclides were considered the most 

important to dose for workers and members of the public (Wruble and McDowell 1990a). In addition, 

since the mid-1960s, measurements have been reported for gross alpha and gross beta 

concentrations (Glora and Brown 1964b). The following sections discuss the history of these 

measurements in the NTS annual environmental reports (see References section) and their 

importance to dose reconstruction for unmonitored employees.

4.2.1.2.1 Annual Inhalation Intakes of Tritium

Atmospheric measurements at NTS began in 1977 with samplers in Areas 5, 10, and 23 (Lantz 

1978a). The sampler near Building 650 (Mercury) was the control station. During that year, the 

samples in Area 10 (near the Sedan Crater) demonstrated the highest HTO concentrations with a 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 17 of 88

Figure 4-1. Air sampling stations in 1997. 

high of 3.0 × 10-4 pCi/m3

 (Lantz 1978a). The number of sampling locations increased over the 

following years to include all the areas in Table 4-1. These locations provided representative samples 

from the most populated areas on the site (Black and Townsend 1998).

The tritium concentrations in Table 4-1 are typically the average of the maximum concentrations for a 

given area in a given year, unless these values were determined to be adversely affected by nearby 

point-source releases (e.g., Building 790 in 1983; Scoggins 1983, p. 29). If maximum values were not 

provided, the average concentrations were reported. 
Document No. ORAUT-TKBS-0008-4 Revision No. 0

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Table 4-1. HTO atmospheric concentrations by year and area with estimated maximum and average annual organ dose.

a

Year

Area

Site 

average

(pCi/mL)

Site 

maximum

(pCi/mL)

Site 

maximum 

annual 

dose

(mrem)

Site 

average 

annual 

dose

(mrem)

1 3 5 6 10 12 15 16 18 20 23 25

1977 3.0E-04 1.2E-06 1.5E-04 3.0E-04 6.9E-02 3.5E-02

1978 2.9E-03 1.3E-04 1.9E-05 1.9E-05 3.0E-05 6.2E-04 2.9E-03 6.7E-01 1.4E-01

1979 6.2E-04b 2.9E-03b 6.7E-01b 1.4E-01b

1980 2.6E-04 7.1E-05 1.7E-04 2.6E-04 6.0E-02 3.8E-02

1981 3.6E-03 9.8E-04 2.3E-03 3.6E-03 8.3E-01 5.3E-01

1982 7.5E-04 2.3E-03 4.4E-03 7.6E-04 5.0E-03 2.6E-03 5.0E-03 1.2E+00 6.1E-01

1983 8.2E-05 1.1E-04 8.9E-05 3.9E-03 6.5E-04 9.6E-04 3.9E-03 8.9E-01 2.2E-01

1984 1.7E-04 5.3E-04 3.3E-05 3.5E-05 5.0E-03 1.2E-03 5.0E-03 1.2E+00 2.7E-01

1985 1.4E-04 8.6E-05 4.7E-03 3.5E-05 4.7E-04 1.1E-03 4.7E-03 1.1E+00 2.5E-01

1986 1.2E-04 5.6E-05 6.6E-05 6.0E-05 8.4E-04 2.3E-04 8.4E-04 1.9E-01 5.3E-02

1987 4.5E-05 1.7E-04 6.3E-05 4.6E-04 2.6E-04 2.0E-04 4.6E-04 1.0E-01 4.6E-02

1988 4.4E-05 3.9E-05 2.0E-05 5.2E-04 3.2E-05 1.3E-04 5.2E-04 1.2E-01 3.0E-02

1989 1.7E-04 3.1E-05 1.3E-05 1.9E-05 8.2E-05 4.6E-06 5.3E-05 1.7E-04 3.9E-02 1.2E-02

1990 7.4E-06 2.2E-05 1.3E-05 4.0E-06 2.3E-05 7.5E-05 2.1E-05 2.4E-05 7.5E-05 1.7E-02 5.5E-03

1991 9.1E-06 2.8E-05 6.3E-06 8.4E-06 1.7E-05 4.3E-06 2.1E-06 1.1E-05 2.8E-05 6.5E-03 2.5E-03

1992 1.0E-05 3.5E-05 3.3E-06 6.2E-05 6.2E-05 6.1E-06 8.8E-05 3.8E-05 8.8E-05 2.0E-02 8.8E-03

1993 8.7E-06 2.8E-05 7.9E-06 3.7E-06 2.0E-05 4.0E-06 1.9E-06 1.1E-05 2.8E-05 6.5E-03 2.4E-03

1994 4.00E-06 4.70E-05 5.10E-06 2.20E-06 2.60E-06 3.70E-05 4.10E-06 2.90E-06 1.06E-05 2.81E-05 6.74E-03 2.55E-03

1995 3.7E-06 1.4E-06 2.1E-05 9.5E-07 7.0E-06 4.3E-06 1.0E-05 2.6E-06 1.4E-06 5.8E-06 2.1E-05 4.9E-03 1.3E-03

1996 3.1E-06 4.1E-06 1.6E-05 5.6E-05 2.6E-05 2.0E-05 1.4E-05 1.1E-06 1.1E-06 1.6E-05 5.6E-05 1.3E-02 3.6E-03

1997 2.2E-06 3.8E-06 7.5E-06 5.6E-05 2.9E-05 2.3E-05 2.0E-05 1.3E-06 1.3E-06 1.6E-05 5.6E-05 1.3E-02 3.7E-03

1998 3.4E-06 1.7E-05 1.8E-04 2.9E-05 5.5E-05 1.4E-05 4.6E-04 1.1E-04 4.6E-04 1.1E-01 2.5E-02

1999 2.0E-05 1.8E-05 4.2E-06 4.1E-05 2.8E-05 2.7E-05 7.5E-04 1.3E-04 7.5E-04 1.7E-01 2.9E-02

2000 4.1E-06 1.5E-05 5.1E-05 5.4E-05 1.9E-05 9.7E-04 1.9E-04 9.7E-04 2.2E-01 4.3E-02

2001 4.2E-06 3.6E-06 6.2E-06 4.2E-05 2.5E-05 3.8E-06 2.3E-06 1.3E-06 1.1E-03 7.1E-07 6.1E-06 1.1E-04 1.1E-03 2.5E-01 2.5E-02

a. Sources: Environmental Reports listed in the References section.

b. Value not provided in annual reports; assumed same as that for 1978.
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 19 of 88

In addition to the annual area tritium concentrations, Table 4-1 lists NTS site average and maximum 

concentrations, which represent the arithmetic averages of the concentrations of all the areas and the 

maximum of all the areas, respectively. For dose reconstruction, these average and maximum site 

concentrations have been converted to annual organ dose using the dose conversion factor for 

inhalation of 1.80 × 10-11 Sv/Bq (ICRP 1998) and by assuming submersion in the respective 

concentrations for 2,000 hr/yr and an inhalation rate of 2,400 m3

/yr. 

In addition to the inhalation dose, a factor of 1.5 was included in the organ dose calculation to account 

for absorption through the skin.

The uncertainty that is associated with the values in Table 4-1 was not provided in the annual 

environmental reports until 2001. Based on the values for that year, the standard deviation would be 

typically less than 10% of the measured values. Because sampling methods were changed over time

and the unavailability of uncertainty estimates for previous years, the assumption that the 50thpercentile expected intakes are those in Table 4-1 and the 95th-percentile values are twice the Table

4-1 values is favorable to claimants. 

For dose reconstruction, the geometric standard deviation (GSD) of the values in Table 4-2 would be 

estimated by using the following expression for a lognormal distribution:

)()( 1.645

1

50th percentile

95th percentile GSD = (4-1)

If the assumption is made that the 95th-percentile values are twice the Table 4-1 values, the GSD can 

be shown using Equation 4-1 to be 1.52.

If the assumption that is favorable to claimants is used to estimate the 95th-percentile dose (i.e., add 

100%) from the maximum annual doses in Table 4-1, the resultant annual organ doses would be less 

than 2 mrem/yr. 

4.2.1.2.2 Annual Intakes of Plutonium

Routine isotopic atmospheric measurements of plutonium at NTS began in 1971 with samplers in 15 

locations across the site (Lantz 1978a). Six additional sampling stations were added in 1978 (Lantz 

1979). Equipment at fixed locations continuously sampled the ambient air to monitor for radioactive 

materials. These locations were chosen to provide representative samples from populated areas on 

the site and to monitor resuspension of low-fired plutonium spread by safety experiments before 1960 

in Areas 2, 3, 4, 7, 9, and 10. Access, worker population, geographical coverage, presence of 

radioactivity, and availability of electric power were considerations in the site selection for air samplers 

(Black and Townsend 1997).

In 1988, efforts to monitor radioactive air emissions at NTS increased as a result of the requirements 

of DOE Order 5400.1 (DOE 1990). Known and potential effluent sources throughout NTS were 

assessed for their potential to contribute to public dose (Black and Townsend 1997).

The 239Pu concentrations in Table 4-2 for 1989 through 2001 represent the average of the maximum 

concentrations for a given area in a given year. For cases in which maximum values were not 

provided (i.e., 1971 through 1988), the average of the average concentrations was listed. In addition 

to annual area concentrations, Table 4-2 lists NTS site average and maximum concentrations, which 

represent the arithmetic averages of the concentration of all the areas and the maximum of the area 

maximum or area averages of all the areas, respectively. Potential intakes associated with these 

concentrations can be calculated under the assumption that an unmonitored worker was 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 20 of 88

occupationally exposed for 2,000 hours per year and had a breathing rate of 2,400 m3/yr. Table 4-3 

lists these calculated intakes.

Some covered employees remained on the site continuously for weeks at a time. However, because 

most nonworking hours were spent indoors where ambient air particulate loadings would be much 

less than outdoor loadings and because of the conservative assumptions that were used to estimate 

the values in Table 4-3, adjustment of the tabular data is not required to ensure intakes are not 

underestimated for these individuals. In addition, employees who lived on the site during their work 

week would have been housed in Area 12 or Area 23 (Mercury). For most years, the values in 

Table 4-2 for these locations are less than the site average values. For dose reconstruction, the 

inhalation intake values in Table 4-3 can be assumed to be bounding; therefore, the organ dose from 

these intakes should be applied with a constant distribution. 

It is assumed that plutonium could be any of absorption type S, Super S, or M depending on which 

type delivers the maximum organ dose. Because these doses are based on air monitoring results, 

evaluation for type Super S (ORAUT 2010b) is required for lung and thoracic cancers if there were no 

bioassay results for the employee.

4.2.1.2.3 Annual Intakes of Other Radionuclides

Extensive studies were performed in the 1980s to quantify residual contamination at NTS (McArthur 

and Kordas 1983, 1985; McArthur and Mead 1987, 1988, 1989; McArthur 1991). Table 4-4 lists the 

results of these studies (McArthur 1991). Table 4-5 lists the total areal depositions based on the 

inventory values in Table 4-4 divided by the areal size. The results in Table 4-5 are representative of 

areas of NTS that contain measurable levels of contamination. These areas actually represent only 

about one-third of the total area within the boundaries of NTS.

In addition, because the data in Table 4-4 are representative of soil contamination in 1991, the values 

in Table 4-5 were decay-corrected to the beginning of 1963 – the first year after the cessation of 

atmospheric testing. 

4.2.1.2.4 Scaling Factors for Inhalation Intakes

Because the air sampling program did not provide isotopic analyses for all the radionuclides identified 

in NTS soils, scaling factors were developed to estimate potential intakes for these radionuclides 

based on their relative abundances in comparison with 239Pu when the soil contamination data 

(McArthur 1991) have been decay-corrected to 1963. These area-specific ratios are listed in 

Table 4-6.

The scaling factors in Table 4-6 are used in conjunction with the 239Pu intakes in Table 4-3 to 

determine potential environmental intakes of all the radionuclides important to dose (in Table 4-6) that 

have been identified as persistent in NTS soils. For dose reconstruction, maximum intakes of these 

radionuclides are calculated by selecting the maximum annual intake of plutonium (e.g., 0.381 Bq/yr 

derived for Area 9 in 1972) and multiplying this value by the maximum scaling factor for each of the 

radionuclides in Table 4-6. 

4.2.1.2.5 Correction for Resuspension for Early Times After Atmospheric Tests

Anspaugh et al. (2002) stated that, based on empirical observations, concentrations of resuspended 

radionuclides in air have been noted to display a strong time dependence early after deposition and 

that this pathway could be important for reoccupation of contaminated property. Anspaugh et al. also 

stated that there has not been universal agreement that resuspension is an important pathway but 

that it is now generally accepted that there are a few instances in which the pathway could be the 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 21 of 88

Table 4-2. Atmospheric concentrations of 239Pu for sampled areas (pCi/m

3

).a 

Year

Area

12345679 10 11 12

1971 3.7E-04 6.6E-04 1.7E-04 1.2E-04 1.7E-04 7.2E-04 2.1E-04 9.1E-05 1.6E-04

1972 1.5E-04 2.2E-04 3.7E-04 1.4E-04 1.3E-03 4.3E-03 2.9E-04 2.4E-04 7.9E-04

1973 8.7E-05 2.2E-04 2.1E-04 8.7E-05 1.2E-04 8.6E-04 4.4E-05 2.4E-04 4.3E-05

1974 8.0E-05 6.8E-05 1.3E-04 5.0E-05 7.9E-05 2.1E-04 5.7E-05 8.0E-05 5.8E-05

1975 2.9E-05 3.8E-05 1.5E-04 4.3E-05 6.3E-05 1.7E-04 5.7E-05 4.9E-05 3.3E-05

1976 2.9E-04 1.7E-04 1.0E-03 1.5E-04 4.4E-04 3.2E-03 2.1E-04 2.4E-04 2.8E-04

1977 2.7E-05 6.5E-05 2.6E-04 3.4E-05 5.5E-05 2.5E-04 4.6E-05 1.2E-04 2.1E-05

1978 4.7E-05 1.3E-04 2.8E-04 6.1E-05 9.9E-05 5.4E-04 2.9E-04 1.1E-04 6.2E-05

1979 7.4E-05 4.7E-04 1.7E-06 2.1E-05 4.0E-05 4.9E-05 5.2E-04 2.9E-04 3.7E-05 2.4E-05

1980 2.3E-05 2.6E-04 1.1E-06 2.9E-05 4.8E-05 4.5E-05 4.5E-04 4.4E-05 1.9E-05

1981 2.6E-05 6.4E-05 1.4E-04 2.5E-05 2.9E-05 3.3E-05 3.2E-04 4.6E-05 2.0E-05

1982 6.1E-05 4.0E-05 6.2E-05 2.0E-05 3.5E-05 7.4E-05 2.2E-04 5.3E-05 2.7E-05

1983 3.1E-05 6.5E-05 9.2E-05 2.3E-05 3.2E-05 3.3E-05 2.1E-04 1.8E-04 2.3E-05

1984 1.9E-05 1.0E-04 2.3E-04 3.8E-05 2.0E-05 3.8E-05 1.0E-03 5.6E-05 1.7E-05

1985 1.5E-05 5.8E-05 2.1E-04 3.1E-05 2.7E-05 2.7E-05 1.3E-03 5.2E-05 2.3E-05

1986 1.5E-04 6.2E-05 5.5E-04 2.8E-05 1.9E-04 4.8E-05 2.8E-04 2.5E-05 3.2E-05

1987 5.5E-05 2.3E-05 2.8E-03 1.7E-05 2.5E-05 1.6E-05 1.1E-04 3.2E-05 1.6E-05

1988 9.2E-05 2.8E-05 2.1E-04 2.8E-05 2.2E-05 1.5E-05 5.1E-05 1.2E-05 1.6E-05

1989 8.6E-04 5.8E-05 5.4E-04 5.0E-05 8.0E-05 2.7E-04 3.5E-04 5.4E-04 6.8E-04

1990 9.9E-05 3.2E-05 3.5E-04 3.1E-05 3.7E-05 4.1E-05 5.0E-04 4.3E-05 1.1E-04 1.0E-05

1991 3.1E-03 2.1E-05 3.8E-04 3.5E-05 7.2E-05 3.2E-05 3.5E-04 5.4E-07 2.2E-04 1.4E-05

1992 1.5E-04 1.1E-04 2.3E-03 8.0E-05 7.3E-05 3.9E-04 8.8E-04 2.4E-04 8.8E-05 3.8E-05

1993 2.0E-04 4.9E-05 4.0E-04 8.0E-05 9.0E-05 7.8E-05 2.7E-03 1.7E-04 4.2E-04 8.4E-04

1994

1995 4.1E-05 1.1E-05 2.4E-04 1.5E-05 5.0E-05 1.2E-05 3.9E-04 3.8E-05 4.0E-05 1.1E-05

1996 7.5E-04 3.0E-04 1.6E-05 4.6E-05 4.5E-04 6.1E-04 5.8E-05 1.9E-05 3.7E-06

1997 1.1E-05 2.6E-05 1.1E-05 1.5E-06

1998 4.7E-04 6.7E-05 9.5E-05 1.1E-05 2.7E-05 7.4E-04 9.5E-06

1999 5.9E-05

2000 8.5E-05 2.7E-04 5.9E-05 2.1E-05 4.1E-04 3.4E-05 2.8E-03

2001 4.5E-04 1.2E-04 1.2E-04 1.9E-04 6.8E-05 9.6E-05 1.7E-05 5.0E-04 8.5E-06
Document No. ORAUT-TKBS-0008-4 Revision No. 0

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Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1971 1.9E-04 8.0E-05 2.3E-04 9.4E-05 8.4E-05 3.9E-05 2.26E-04 7.21E-04

1972 1.7E-04 1.5E-04 2.7E-03 3.5E-04 1.7E-04 5.1E-05 7.63E-04 4.29E-03

1973 4.4E-05 3.8E-05 7.2E-05 5.6E-05 3.7E-05 3.3E-05 1.46E-04 8.59E-04

1974 6.0E-05 6.6E-05 8.5E-05 8.6E-05 7.2E-05 1.2E-04 8.64E-05 2.11E-04

1975 4.3E-05 3.4E-05 4.4E-05 5.7E-05 6.8E-05 7.0E-05 6.27E-05 1.68E-04

1976 1.6E-04 9.9E-05 5.7E-04 2.0E-04 1.4E-04 6.3E-05 4.82E-04 3.18E-03

1977 2.6E-05 2.3E-05 2.7E-05 3.3E-05 3.1E-05 2.3E-05 6.93E-05 2.55E-04

1978 5.1E-05 6.2E-05 5.8E-05 7.0E-05 9.4E-05 3.9E-05 1.33E-04 5.42E-04

1979 5.2E-05 2.1E-05 2.0E-05 2.2E-05 1.8E-05 1.6E-05 2.3E-05 9.98E-05 5.24E-04

1980 6.4E-05 9.5E-05 3.5E-05 3.1E-05 2.1E-05 1.6E-05 7.88E-05 4.52E-04

1981 9.3E-05 2.6E-05 2.3E-05 2.9E-05 2.4E-05 2.5E-05 1.6E-05 5.88E-05 3.22E-04

1982 4.2E-05 1.7E-05 2.2E-05 2.1E-05 1.7E-05 5.04E-05 2.15E-04

1983 2.2E-05 1.5E-05 3.0E-05 2.8E-05 3.0E-05 5.85E-05 2.14E-04

1984 3.2E-05 2.8E-05 2.5E-05 2.9E-05 3.2E-05 4.8E-05 1.15E-04 1.02E-03

1985 4.0E-05 2.1E-05 3.5E-05 2.2E-05 2.6E-05 3.1E-05 1.30E-04 1.33E-03

1986 3.9E-05 2.8E-05 2.2E-05 6.6E-05 2.1E-05 2.3E-05 1.03E-04 5.45E-04

1987 1.8E-05 1.4E-05 1.7E-05 1.6E-05 2.3E-05 2.13E-04 2.81E-03

1988 2.4E-05 1.4E-05 3.5E-05 1.5E-05 1.5E-05 4.15E-05 2.14E-04

1989 2.3E-05 1.1E-05 9.0E-06 1.2E-04 9.5E-06 4.8E-06 2.40E-04 8.55E-04

1990 8.2E-05 5.8E-06 3.9E-06 8.5E-05 1.1E-05 3.5E-05 9.17E-05 4.98E-04

1991 1.2E-04 2.9E-05 8.8E-06 1.3E-05 2.6E-05 7.7E-06 2.77E-04 3.10E-03

1992 1.1E-03 1.9E-05 8.9E-06 1.3E-05 1.8E-05 1.6E-04 3.52E-04 2.26E-03

1993 3.3E-04 1.3E-05 7.7E-06 6.3E-06 3.4E-06 1.1E-05 3.37E-04 2.70E-03

1994

1995 1.6E-04 1.4E-05 1.0E-06 7.4E-06 4.5E-06 6.92E-05 3.90E-04

1996 2.4E-04 1.8E-06 3.5E-06 4.5E-06 6.3E-06 4.2E-06 2.6E-06 1.57E-04 7.50E-04

1997 2.9E-05 1.9E-07 2.0E-05 2.0E-05 6.0E-07 1.0E-06 1.6E-06 1.11E-05 2.90E-05

1998 6.3E-05 3.1E-05 3.3E-05 1.2E-05 6.3E-06 1.30E-04 7.35E-04

1999 1.1E-05 1.1E-05 2.0E-05 6.2E-06 2.14E-05 5.90E-05

2000 2.6E-04 1.0E-05 6.6E-06 1.1E-05 3.63E-04 2.83E-03

2001 1.4E-05 9.6E-06 1.7E-05 7.6E-06 2.0E-06 4.4E-06 1.02E-04 5.04E-04

a. Sources: Environmental Reports listed in the References section.

Table 4-3. Annual inhalation intakes from 239Pu for sampled areas (Bq) (from Attachment A).

Year

Area

1 2 3 4 5 6 7 9 10 11 12

1971 0.0329 0.0587 0.0147 0.0105 0.0153 0.0640 0.0185 0.0081 0.0145

1972 0.0131 0.0199 0.0327 0.0124 0.1190 0.3810 0.0258 0.0216 0.0698

1973 0.0077 0.0198 0.0183 0.0078 0.0107 0.0763 0.0039 0.0216 0.0038
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 23 of 88

Year

Area

12345679 10 11 12

1974 0.0071 0.0060 0.0112 0.0044 0.0070 0.0187 0.0051 0.0071 0.0052

1975 0.0026 0.0034 0.0130 0.0038 0.0056 0.0149 0.0051 0.0044 0.0029

1976 0.0255 0.0153 0.0924 0.0133 0.0392 0.2824 0.0189 0.0210 0.0248

1977 0.0024 0.0058 0.0226 0.0030 0.0049 0.0225 0.0041 0.0107 0.0019

1978 0.0042 0.0118 0.0250 0.0054 0.0088 0.0481 0.0258 0.0095 0.0055

1979 0.0066 0.0413 0.0001 0.0018 0.0035 0.0044 0.0465 0.0258 0.0033 0.0021

1980 0.0020 0.0230 0.0001 0.0026 0.0043 0.0040 0.0401 0.0039 0.0017

1981 0.0023 0.0056 0.0124 0.0022 0.0026 0.0029 0.0286 0.0041 0.0018

1982 0.0054 0.0036 0.0055 0.0018 0.0031 0.0066 0.0191 0.0047 0.0024

1983 0.0028 0.0057 0.0082 0.0020 0.0029 0.0029 0.0190 0.0162 0.0020

1984 0.0017 0.0088 0.0202 0.0034 0.0018 0.0034 0.0906 0.0050 0.0015

1985 0.0013 0.0051 0.0189 0.0027 0.0024 0.0024 0.1181 0.0046 0.0020

1986 0.0131 0.0055 0.0484 0.0025 0.0165 0.0043 0.0249 0.0022 0.0028

1987 0.0048 0.0020 0.2495 0.0015 0.0022 0.0014 0.0098 0.0028 0.0014

1988 0.0082 0.0025 0.0190 0.0025 0.0020 0.0013 0.0045 0.0011 0.0014

1989 0.0759 0.0052 0.0478 0.0045 0.0071 0.0240 0.0311 0.0480 0.0604

1990 0.0088 0.0029 0.0308 0.0027 0.0033 0.0036 0.0442 0.0038 0.0095 0.0009

1991 0.2753 0.0018 0.0340 0.0031 0.0064 0.0029 0.0311 0.0000 0.0199 0.0013

1992 0.0133 0.0098 0.2007 0.0071 0.0065 0.0346 0.0781 0.0213 0.0078 0.0034

1993 0.0180 0.0044 0.0352 0.0071 0.0079 0.0069 0.2398 0.0151 0.0373 0.0746

1994

1995 0.0036 0.0010 0.0216 0.0014 0.0045 0.0011 0.0346 0.0033 0.0036 0.0010

1996 0.0666 0.0000 0.0268 0.0015 0.0041 0.0400 0.0542 0.0051 0.0017 0.0003

1997 0.0010 0.0023 0.0010 0.0001

1998 0.0416 0.0060 0.0084 0.0010 0.0024 0.0653 0.0008

1999

2000 0.0075 0.0239 0.0052 0.0018 0.0366 0.0030 0.2509

2001 0.0404 0.0011 0.0103 0.0166 0.0061 0.0086 0.0015 0.0447 0.0008
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 24 of 88

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1971 0.0171 0.0071 0.0201 0.0083 0.0075 0.0035 0.0200 0.0640

1972 0.0150 0.0129 0.2424 0.0309 0.0147 0.0045 0.0677 0.3810

1973 0.0039 0.0034 0.0063 0.0050 0.0033 0.0029 0.0130 0.0763

1974 0.0053 0.0059 0.0075 0.0076 0.0064 0.0104 0.0077 0.0187

1975 0.0038 0.0030 0.0039 0.0051 0.0060 0.0062 0.0056 0.0149

1976 0.0139 0.0088 0.0504 0.0178 0.0122 0.0056 0.0428 0.2824

1977 0.0023 0.0020 0.0024 0.0029 0.0028 0.0020 0.0062 0.0226

1978 0.0045 0.0055 0.0052 0.0062 0.0083 0.0035 0.0118 0.0481

1979 0.0046 0.0019 0.0017 0.0020 0.0016 0.0014 0.0020 0.0089 0.0465

1980 0.0057 0.0085 0.0031 0.0027 0.0019 0.0014 0.0070 0.0401

1981 0.0083 0.0023 0.0020 0.0025 0.0022 0.0022 0.0014 0.0052 0.0286

1982 0.0037 0.0015 0.0019 0.0019 0.0015 0.0045 0.0191

1983 0.0020 0.0013 0.0026 0.0025 0.0027 0.0052 0.0190

1984 0.0028 0.0025 0.0022 0.0025 0.0028 0.0043 0.0102 0.0906

1985 0.0036 0.0019 0.0031 0.0020 0.0023 0.0028 0.0115 0.1181

1986 0.0035 0.0025 0.0020 0.0058 0.0018 0.0020 0.0092 0.0484

1987 0.0016 0.0012 0.0015 0.0014 0.0020 0.0000 0.0189 0.2495

1988 0.0022 0.0012 0.0031 0.0013 0.0013 0.0037 0.0190

1989 0.0021 0.0010 0.0008 0.0104 0.0008 0.0004 0.0213 0.0759

1990 0.0073 0.0005 0.0003 0.0075 0.0010 0.0031 0.0081 0.0442

1991 0.0109 0.0026 0.0008 0.0012 0.0023 0.0007 0.0246 0.2753

1992 0.0977 0.0017 0.0008 0.0011 0.0016 0.0142 0.0312 0.2007

1993 0.0293 0.0012 0.0007 0.0006 0.0003 0.0010 0.0300 0.2398

1994 0.0000 0.0000

1995 0.0142 0.0012 0.0001 0.0007 0.0004 0.0061 0.0346

1996 0.0213 0.0002 0.0003 0.0000 0.0004 0.0006 0.0004 0.0002 0.0124 0.0666

1997 0.0026 0.0000 0.0018 0.0018 0.0001 0.0001 0.0001 0.0010 0.0026

1998 0.0056 0.0027 0.0029 0.0010 0.0006 0.0115 0.0653

1999 0.0010 0.0010 0.0018 0.0006 0.0011 0.0018

2000 0.0229 0.0009 0.0006 0.0010 0.0322 0.2509

2001 0.0012 0.0008 0.0015 0.0007 0.0002 0.0004 0.0090 0.0447
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 25 of 88

Table 4-4. Inventory of contaminated soil (Ci). 

Area Area (mi2) Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 26.5 4.2 6.5 24 1.1 8.8 15 15 0.1 0.5

2 19.7 2.9 8.6 22 1.2 24 46 14 0.4

3 32.3 4.6 3.1 37 1 12 33 18 0.1 0.5

4 16 6.6 13 40 1.6 12 13 9.1 0.2

5 2.9 0.6 0.1 4.8 0.6 0.4 0.9 10 0.2 0

6 32.3 1.7 3.3 8.4 0.2 2.8 3.5 0

7 19.3 2.2 0.6 16 1 5.2 9.2 22 0.2 0.3

8 13.9 17 8 110 5.7 42 25 4.4 0.6

9 20 4.2 2.2 89 0.7 8.7 13 23 0.2 0.3

10 20 19 19 110 9.7 84 55 2.2 0.3 5

11 4 3.3 0.5 29 0 0.5 0.3

12 39.6 5.7 8.5 39 1.2 20 17

15 35.3 8 7.8 63 0.3 19 22

16 14.3 0.7 1.5 3.7 0.1 2.9 3.7

17 31.4 2.8 4.5 18 1.0 15 19

18 27.3 19 5.6 100 0.7 10 17 1.1 0.1 0.8

19 148.3 21 32 140 1.1 36 31

20 6.2 23 30 41 7.9 5.5 4.3 13 1.6 4.8

25 0.9 0.2 0.1 0.4

26 0.2

30 0.03 3.2 4.5 14 0.8 1.5 1.3 0.7 0.1 0.2

Source: McArthur (1991).

Table 4-5. Radionuclide areal soil deposition decay, corrected to 1963 (Bq/m2) 

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 2.37E+03 4.37E+03 1.29E+04 2.35E+04 9.06E+03 1.57E+04 3.47E+04 4.89E+02 1.35E+04

2 2.20E+03 7.78E+03 1.60E+04 3.45E+04 3.32E+04 6.49E+04 4.35E+04 0.00E+00 1.45E+04

3 2.13E+03 1.71E+03 1.64E+04 1.76E+04 1.01E+04 2.84E+04 3.41E+04 4.01E+02 1.11E+04

4 6.16E+03 1.45E+04 3.57E+04 5.67E+04 2.05E+04 2.26E+04 3.48E+04 8.93E+03

5 3.09E+03 6.15E+02 2.37E+04 1.17E+05 3.76E+03 8.63E+03 2.11E+05 8.94E+03 0.00E+00

6 7.86E+02 1.82E+03 3.72E+03 3.51E+03 2.36E+03 3.01E+03

7 1.70E+03 5.54E+02 1.19E+04 2.94E+04 7.35E+03 1.33E+04 6.98E+04 1.34E+03 1.11E+04

8 1.83E+04 1.03E+04 1.13E+05 2.33E+05 8.24E+04 5.00E+04 1.94E+04 3.08E+04

9 3.14E+03 1.96E+03 6.36E+04 1.98E+04 1.19E+04 1.81E+04 7.04E+04 1.30E+03 1.07E+04

10 1.42E+04 1.69E+04 7.86E+04 2.75E+05 1.15E+05 7.65E+04 6.74E+03 1.94E+03 1.79E+05

11 1.23E+04 2.23E+03 1.04E+05 0.00E+00 3.41E+03 2.09E+03

12 2.15E+03 3.83E+03 1.41E+04 1.72E+04 1.38E+04 1.19E+04

15 3.39E+03 3.94E+03 2.55E+04 4.82E+03 1.47E+04 1.73E+04

16 7.31E+02 1.87E+03 3.70E+03 3.97E+03 5.53E+03 7.20E+03

17 1.33E+03 2.55E+03 8.20E+03 1.81E+04 1.30E+04 1.68E+04

18 1.04E+04 3.66E+03 5.24E+04 1.45E+04 9.99E+03 1.73E+04 2.47E+03 4.75E+02 2.09E+04

19 2.12E+03 3.85E+03 1.35E+04 4.21E+03 6.62E+03 5.81E+03

20 5.54E+04 8.62E+04 9.45E+04 7.23E+05 2.42E+04 1.93E+04 1.28E+05 3.34E+04 5.53E+05

25 6.06E+03 3.09E+03 2.72E+04

26

30 1.59E+05 2.67E+05 6.67E+05 1.51E+06 1.36E+05 1.21E+05 1.43E+05 4.32E+04 4.76E+05

dominant one. Many observations have shown that the rate of resuspension decreases rapidly with 

time and, for accident situations, resuspension is only of importance (in comparison with the inhalation 

exposure from the initial cloud passage) over short periods. For this reason, Anspaugh et al. stated 

that the resuspension factor model has been widely used to predict the concentration of resuspended 

radionuclides early after initial deposition while the mass loading model (which uses measurements of 

dust loading in air and soil contamination data to predict air concentration of radionuclides) has 

generally been preferred for times long after deposition. However, Anspaugh et al. also stated that it 

is always preferable to rely on actual measurements that are performed over long periods (such as 

those in Section 4.2.1.2.2).
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 26 of 88

Table 4-6. Abundance of radionuclides in NTS soils in relation to 239Pu decay, corrected to 1963. 

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 0.183 0.338 1.000 1.818 0.700 1.216 2.677 0.038 1.041

2 0.138 0.487 1.000 2.164 2.081 4.067 2.726 0.908

3 0.130 0.104 1.000 1.072 0.619 1.735 2.084 0.024 0.675

4 0.172 0.405 1.000 1.587 0.572 0.632 0.975 0.250

5 0.131 0.026 1.000 4.958 0.159 0.365 8.925 0.378

6 0.212 0.490 1.000 0.944 0.636 0.810

7 0.144 0.047 1.000 2.479 0.620 1.118 5.890 0.113 0.937

8 0.162 0.091 1.000 2.055 0.728 0.442 0.171 0.273

9 0.049 0.031 1.000 0.312 0.187 0.284 1.107 0.020 0.168

10 0.181 0.215 1.000 3.498 1.457 0.973 0.086 0.025 2.271

11 0.119 0.021 1.000 0.033 0.020

12 0.153 0.272 1.000 1.220 0.978 0.848

15 0.133 0.154 1.000 0.189 0.575 0.679

16 0.198 0.505 1.000 1.072 1.495 1.945

17 0.163 0.312 1.000 2.204 1.590 2.053

18 0.199 0.070 1.000 0.278 0.191 0.331 0.047 0.009 0.400

19 0.157 0.285 1.000 0.312 0.491 0.431

20 0.586 0.912 1.000 7.643 0.256 0.204 1.358 0.354 5.849

25

26

30 0.229 0.321 1.000 0.057 0.107 0.093 0.050 0.007 0.014

Maximum scaling factor 0.586 0.912 1.000 7.64 2.08 4.07 8.93 0.378 5.85

Scaled maximum intake 

(Bq/yr)

0.223 0.347 0.381 2.91 0.792 1.55 3.40 0.144 2.23

Anspaugh et al. (2002) presented several resuspension models that have been proposed but 

concluded that they can be over- or under-predictive at various times after deposition in comparison 

with empirical observations. However, with expanded data sets from the work of Hicks (1981e) and 

others in the 1980s, Anspaugh et al. proposed a resuspension model that more accurately describes 

the observed results over the entire timespan of the expanded data set for NTS:

5 0.07 9 0.003 9 1 1 [10 6 10 10 ] 10 m t t Se e f

− − −− − ± − = +× + × (4-2)

A graphical depiction of Equation 4-2 is in Figure 4-2 from time t equal zero to 1,000 days after 

detonation. As shown in Figure 4-2, the resuspension factor Sf ranges from about 10-5 at times early 

after deposition, falls rapidly during the first 100 or so days to a value of about 10-8, and then 

approaches a value of 10-9 after a few years. The factor of 10 at the end of the equation is a 

statement of uncertainty in the model. 

If it is true that the mass loading approach is more predictive at time long after initial deposition and 

that the resuspension proposed by Anspaugh et al. is predictive of the observed results over the 

expanded dataset (including those developed in the 1980s), the factor of 10-9 could be taken to be the 

resuspension factor that would be predictive of the mass loading process that is thought to be more 

important during the times when air monitoring data are available (i.e., 1971 through 2001, see 

Section 4.2.1.2.2). 

When Anspaugh et al.’s proposed resuspension model (Equation 4-2) is integrated over 180 to 

545 days and compared with the result of the integral of the constant 10-9 over the same period, a 

factor is developed that can be used to correct the intakes derived from air sampling data (i.e., times 

long after initial deposition) for the early resuspension phenomenon that has been observed at NTS. 

The period of integration was selected to begin at day 180 because the last atmospheric tests at NTS 

were in July 1962 and, therefore, about 180 days had passed before the beginning of 1963. 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 27 of 88

Figure 4-2. Resuspension factor as a function of time after initial deposition 

(from Attachment A). 

For the Anspaugh et al. (2002) model, the early resuspension correction factor has been determined 

to be 3.12. In a similar manner, correction factors for 1964 and 1965 were determined to be 1.72 and 

1.24, respectively. Thus, for dose reconstruction, the scaled maximum inhalation intakes in Table 4-6 

should be multiplied by these factors to account for early resuspension from 1963 through 1965. If 

necessary, these intakes can be prorated for time less than a year for a best estimate if the worker 

was on the site for only a fraction of the year.

4.2.1.2.6 Correction for Inhalation Dose from Short-Lived Fission and Activation Products

Correction factors were developed to account for inhalation intakes of short-lived fission and activation 

products based on organ-specific dose from the 90Sr intakes in Table 4-7 (from Attachment A). These 

corrections are listed in Table 4-8.

Table 4-7. Scaled inhalation intakes corrected for early resuspension (Bq/yr). 

Year of intake Am-241 Pu-238 Pu-239, 240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1963 0.70 1.08 1.19 9.08 2.47 4.84 10.61 0.45 6.96

1964 0.38 0.59 0.65 4.98 1.35 2.65 5.81 0.25 3.81

1965 0.28 0.43 0.47 3.61 0.98 1.92 4.22 0.18 2.77

All subsequent years 0.223 0.347 0.381 2.91 0.792 1.55 3.40 0.144 2.23

The Integrated Modules for Bioassay Analysis (IMBA) computer program was used to determine the 

organ doses resulting from the scaled intakes of 90Sr in Table 4-7 for a period of 10 years (i.e., 

4.84 Bq for 1963, 2.65 Bq for 1964, 1.92 for 1965, and 1.55 for all subsequent years through 1972). 

These doses were then multiplied by the correction factors in Table 4-8 to determine the additional 

dose that should be added to account for potential dose from inhalation of short-lived fission and 

activation products. The organ-specific fission and activation doses greater than 0.001 rem are listed 

in Table 4-9. Other organs are not listed in Table 4-9 because all were less than 0.001 rem. These 

doses should be added to the input for the Interactive RadioEpidemiological Program (IREP) as a 

constant as 100% 30-to-250-keV photons. These doses are input as constants because they are 

Resuspension Factor

1.3 E-09 m-1

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

0 100 200 300 400 500 600 700 800 900 1000

Days

meter-1
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 28 of 88

based on bounding intakes (i.e., maximum) and they are entered as 30-to-250-keV photons as an 

assumption favorable to claimants.

Table 4-8. Organ-specific inhalation dose fission and activation product correction factors. 

Organ

Fission and activation product correction factor by year

1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Skin Adrenals

Thymus SI 

Spleen Skin 

Muscle Uterus

Pancreas Kidneys

Breast Testes

Esophagus Ovaries

Brain Stomach

Thyroid Gall bladder

730 364 242 182 145 121 104 90.7 80.6 72.5

ULI 458 179 99.2 64.0 45.0 33.4 25.9 20.6 16.8 14.0

Urinary bladder 335 149 91.3 63.6 47.5 37.2 30.1 24.9 21.0 18.0

Lungs 34,900 14,200 7,960 5,150 3,630 2,700 2,100 1,660 1,360 1,130

ET ET1 ET2 LN(TH)

LN(ET)

1,570 827 598 492 438 412 412 412 412 412

LLI 420 142 70.8 42.4 28.2 20.1 15.1 11.7 9.4 7.6

Colon 390 148 79.4 50.0 34.5 25.2 19.3 15.2 12.3 10.2

Liver 9,260 4,620 1,540 1,190 988 858 769 706 661 629

Red bone marrow 37.9 18.2 12.8 10.4 10.4 10.4 10.4 10.4 10.4 10.4

Bone surfaces 78.5 40.1 28.1 22.4 22.4 22.4 22.4 22.4 22.4 22.4

Table 4-9. Inhalation dose from short-lived fission and activation products (rem). 

Organ

Year

1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Liver 0.002

Bone surface 0.001 0.002 0.002 0.002 0.002 0.002 0.002 0.002 0.002 0.002

LLI 0.001

ET 0.002

Lung 0.008 0.003 0.002 0.001

ET1 1.635 0.474 0.248 0.164 0.146 0.138 0.138 0.138 0.138 0.138

4.2.2 Annual Intakes from Ingestion

To account for potential intakes from inadvertent ingestion of contaminated soil, the area-specific 

radionuclide soil deposition data in Table 4-5 were converted to volumetric data (i.e., Bq/g) by 

assuming a radionuclide relaxation depth of 2.3 cm and a soil density of 1.5 g/cm3 (DOE 2003). The 

area-specific radionuclide soil concentrations are listed in Table 4-10.

If the assumption is made that the workers ingested 100 mg of soil each day [EPA (1989) 

recommends a value of 50 mg/d] and that full-time employment was 250 d/yr, annual ingestion can be 

calculated in a manner favorable to claimants. The area-specific annual ingestion rates are in Table

4-11. 

For most radionuclides, Area 30 provided the highest areal deposition and resultant intakes. Area 30 

is relatively small (150 km2

) and inaccessible, and is on the Western edge of NTS. It has rugged 

terrain and includes the northern reaches of Fortymile Canyon. In 1968, it was the site of Project 

BUGGY, the first nuclear row-charge experiment in the PLOWSHARE Program. As a result of the 

test, a trench 255 m long, 77 m wide, and 206 m deep was created. The test resulted in large 

quantities of vitrified glass. Because of the bias that is introduced when Area 30 is included, the 

maximum annual intakes in Table 4-11 have been provided without the Area 30 areal concentrations.

As with inhalation intakes, a method was developed to adjust ingestion doses for potential dose from 

short-lived fission and activation products that are no longer persistent in NTS soils in measurable 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 29 of 88

amounts. The organ-specific fission and activation product correction factors were developed based

on the relative contribution of 90Sr to the total ingestion dose. These correction factors are listed in 

Table 4-12. 

Table 4-10. Radionuclide soil concentration by area (Bq/g). 

Area Am-241 Pu-238 Pu-239, 240 Co-60 Cs-137 Sr-90 Eu-152 Eu-154 Eu-155

1 0.069 0.127 0.375 0.682 0.263 0.456 1.005 0.014 0.391

2 0.064 0.226 0.463 1.001 0.963 1.882 1.262 0.420

3 0.062 0.050 0.475 0.509 0.294 0.824 0.989 0.012 0.321

4 0.179 0.420 1.036 1.644 0.593 0.655 1.010 0.259

5 0.090 0.018 0.686 3.401 0.109 0.250 6.122 0.259

6 0.023 0.053 0.108 0.102 0.069 0.087 0.000 0.00

7 0.049 0.016 0.344 0.852 0.213 0.384 2.024 0.039 0.322

8 0.530 0.297 3.280 6.741 2.389 1.450 0.562 0.894

9 0.091 0.057 1.844 0.575 0.344 0.524 2.042 0.038 0.311

10 0.411 0.491 2.279 7.973 3.321 2.217 0.195 0.056 5.177

11 0.357 0.065 3.005 0.099 0.060

12 0.062 0.111 0.408 0.498 0.399 0.346

15 0.098 0.114 0.740 0.140 0.426 0.502

16 0.021 0.054 0.107 0.115 0.160 0.209

17 0.039 0.074 0.238 0.524 0.378 0.488

18 0.301 0.106 1.518 0.421 0.290 0.502 0.072 0.014 0.607

19 0.061 0.111 0.391 0.122 0.192 0.169

20 1.607 2.500 2.741 20.946 0.701 0.559 3.723 0.969 16.031

25 0.176 0.090 0.789

26

30 4.620 7.749 19.340 43.835 3.953 3.493 4.143 1.252 13.804

Table 4-11. Area-specific and maximum annual ingestion rates (Bq/yr). 

Area

Am-241 Pu-238

Pu-239, 

240 Co-60 Cs-137 Sr-90 Eu-152

Eu154 Eu-155

1 1.72 3.17 9.38 17.06 6.56 11.41 25.12 0.35 9.77

2 1.59 5.64 11.57 25.03 24.08 47.06 31.54 10.51

3 1.54 1.24 11.87 12.72 7.34 20.59 24.73 0.29 8.01

4 4.47 10.49 25.90 41.10 14.83 16.37 25.24 6.47

5 2.24 0.45 17.15 85.03 2.73 6.25 153.05 6.48 0.00

6 0.57 1.32 2.69 2.54 1.71 2.18

7 1.23 0.40 8.59 21.29 5.33 9.61 50.59 0.97 8.05

8 13.24 7.43 81.99 168.52 59.73 36.25 14.05 0.00 22.34

9 2.27 1.42 46.11 14.38 8.60 13.10 51.04 0.94 7.76

10 10.29 12.27 56.98 199.31 83.02 55.42 4.88 1.41 129.41

11 8.93 1.61 75.12 0.00 2.47 1.51

12 1.56 2.77 10.20 12.45 9.98 8.65

15 2.45 2.85 18.49 3.49 10.64 12.56

16 0.53 1.35 2.68 2.87 4.01 5.21

17 0.97 1.85 5.94 13.09 9.44 12.19

18 7.54 2.65 37.95 10.54 7.24 12.55 1.79 0.34 15.17

19 1.53 2.79 9.78 3.05 4.80 4.21

20 40.17 62.49 68.51 523.64 17.54 13.98 93.06 24.23 400.77

25 4.39 2.24 19.73

26

30 115.50 193.72 483.50 1,095.89 98.83 87.33 103.56 31.30 345.10

Maxa 40.17 62.49 81.99 523.64 83.02 55.42 153.05 24.23 400.77

a. Maximum value with Area 30 excluded.

To simplify the application of organ-specific ingestion dose from short-lived fission and activation 

products, the IMBA computer program was used to determine organ-specific annual doses for the 90Sr 

intake of 55.42 Bq/yr for 1963 through 1972. These doses are listed in Table 4-13. The doses in 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 30 of 88

Table 4-13 are multiplied by the Table 4-12 organ-specific correction factors to provide the annual

doses resulting from ingestion of short-lived fission and activation products. The annual organspecific doses that were greater than 0.001 rem are listed in Table 4-14. 

Table 4-12. Organ-specific ingestion fission and activation correction factors (from Attachment A). 

Organ

Correction factor by year

1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals, breast 

Brain, skin, bladder

Stomach, kidneys, muscle, 

pancreas, brain, 

esophagus, SI, liver, 

ovaries ET ET1

ET2 LM(ET) LN(TH), lungs, 

skin, spleen, testes, 

thymus, thyroid, uterus

Gall bladder

416 219 155 123 106 95.0 88.1 84.0 82.0 81.8

ULI 514 184 95.2 58.2 39.3 18.6 16.3 14.6 13.4 12.4

Bone surface/RBM 3.8 3.1 2.7 2.5 2.3 2.3 2.3 2.3 2.3 2.3

LLI/colon 417 208 138 25.6 20.4 17.4 14.6 12.8 11.3 10.2

Table 4-13. Organ-specific annual ingestion doses (rem) for the 90Sr intake of 55.42 Bq/yr. Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Urinary 

bladder

5.68E-06 6.42E-06 6.67E-06 6.87E-06 7.04E-06 7.18E-06 7.31E-06 7.41E-06 7.50E-06 7.58E-06

Bone surface 1.58E-04 3.74E-04 5.47E-04 6.98E-04 8.32E-04 9.50E-04 1.06E-03 1.15E-03 1.23E-03 1.31E-03

Brain 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Breast 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Esophagus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Stomach 3.42E-06 3.78E-06 3.95E-06 4.09E-06 4.21E-06 4.31E-06 4.40E-06 4.47E-06 4.54E-06 4.59E-06

Small intestine 4.71E-06 5.07E-06 5.24E-06 5.38E-06 5.50E-06 5.60E-06 5.69E-06 5.76E-06 5.83E-06 5.88E-06

Upper large 

intestine

3.42E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05

Lower large 

intestine

1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04

Colon 4.36E-03 2.26E-03 1.51E-03 1.14E-03 9.16E-04 7.71E-04 6.61E-04 5.84E-04 5.21E-04 4.71E-04

Kidneys 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Liver 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Muscle 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Ovaries 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Pancreas 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Red marrow 1.08E-04 2.66E-04 3.87E-04 4.86E-04 5.69E-04 6.37E-04 6.94E-04 7.42E-04 7.81E-04 8.14E-04

Extrathoracic 

airways

2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Lungs 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Skin 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Spleen 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Testes 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Thymus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Thyroid 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Uterus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Table 4-14. Organ-specific doses (rem) from ingestion of fission and activation products. 

Organ

Year

1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Bladder 0.002 0.002 0.001

Bone surface 0.001 0.002 0.002 0.002 0.003 0.003 0.003 0.003 0.003

Stomach 0.001

SI 0.002 0.001

ULI 0.015 0.006 0.003 0.002 0.001

LLI 0.048 0.025 0.017 0.003 0.003 0.002 0.002 0.002 0.001 0.001
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 31 of 88

Organ

Year

1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Colon 0.028 0.015 0.010 0.002 0.002 0.001 0.001 0.001

4.3 OCCUPATIONAL EXTERNAL DOSE

Workers incur external doses from ambient radiation levels and noble gases. Ambient radiation 

measurements were reported in NTS annual environmental reports (see References section) starting 

in 1967; no ambient radiation data were provided in the reports for 1968 to 1976. 

4.3.1 Ambient Radiation

Before 1967, ambient radiation levels that were unaffected by weapons testing were not reported in 

the annual environmental reports. Although there were many radiation measurements between 1951 

and 1967, most of these were to characterize the effects of weapons tests and, therefore, were not 

appropriate for use in estimating external environmental dose for unmonitored employees.

In 1967, ambient radiation levels were measured using Victoreen Model 239 indirect reading 

ionization chambers with effective ranges of 0 to 10 mR/hr. Five of these chambers were in small, 

semiprotective enclosures in NTS living areas (REECo 1968). As a backup for the ionization 

chambers, standard NTS film dosimeters were included at each sample location. However, all results 

from these dosimeters for 1967 were either zero or lost due to light or heat damage. Results from 

environmental NTS film dosimeters were subsequently discarded because of the likelihood of heat 

damage.

The ionization chambers were collected on a weekly basis and read on a Victoreen Minometer II 

reader. Corrections were made for background and for nonradiation-induced drift. Two sets of 

chambers were used, one at the sample locations for measurements and another stored fully charged 

in the laboratory. Each week these sets were exchanged, the fresh set being recharged and a record 

kept of the amount of "drift" while stored in the laboratory. A specially designed shock-proof box for 

transportation minimized accidental discharge from mechanical shock.

Readings from the five chambers in each location were averaged to obtain a mean value for each 

location each week. Readings significantly higher than others at a particular location were not used in 

compiling the data because the abnormal readings were most likely to be the result of shock or other 

malfunction and were not representative measurements. Because of the method of background 

subtraction discussed above, the data in the 1967 annual report were assumed to be due to 

manmade radiation in excess of natural background.

No results were reported between 1968 and 1976 in the NTS annual environmental reports; ambient 

radiation reporting was reestablished in 1977 using thermoluminescent dosimeters (TLDs) (Lantz 

1978a). The dosimeters were CaF2:Dy (TLD-200) 0.25- by 0.25- by 0.035-in. chips from Harshaw 

Chemical Company. A badge that consisted of at least two chips shielded by 0.047 in. of cadmium 

(1,030 mg/cm2

) inside a 0.050-in. black plastic (140 mg/cm2) holder was placed about 1 m above the 

ground at each of 10 locations that coincided with air sampling stations. These sites were selected 

because of their proximity to workers. During that year, the natural background ambient radiation 

level was established at 0.26 mrem/d or about 95 mrem/yr. 

In 1987, the Harshaw dosimeters were replaced with Panasonic Model UD-814 TLDs (Gonzalez 

1988). These specifically designed environmental dosimeters contain three identical CaSO4:Tm 

elements and one Li2B4O7:Cu element. The lithium element is shielded with 14 mg/cm2 of material to 

monitor beta particles in the environment. The three calcium elements are encapsulated in 

1,000 mg/cm2 of plastic and lead to monitor ambient gamma levels. The UD-814 TLDs have 

remained in use to the present time.
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 32 of 88

In subsequent years, the number of sampling locations increased to more than 150 and eventually 

covered all populated areas at NTS. Table 4-15 lists average ambient background radiation levels. 

Table 4-16 lists the annual measured ambient radiation including background and site average and 

maximum values. For dose reconstruction, because the values in Table 4-16 represent continuous 

exposure for an entire year, these values should be adjusted for occupational exposure (i.e., 

2,000 hr/yr) and added to the dose of record for unmonitored employees. If the area in which the 

employee worked is known, the average value for that area should be used. If the area is not known, 

the site maximum value should be used to be favorable to claimants.

Table 4-15. Average ambient background radiations levels by year (mrem/yr).

a

Year Level Year Level Year Level

1964 (b) 1980 99 1997 95

1965 (b) 1981 110 1998 88

1966 (b) 1982 100 1999 91

1967 (b) 1983 100 2000 115

1968 (b) 1984 100 2001 114

1969 (b) 1985 100 2002 105

1970 (b) 1986 72 2003 112

1971 (b) 1987 102 2004 110

1972 (b) 1988 131 2005 119

1973 (b) 1989 106 2006 125

1974 (b) 1990 110 2007 121

1975 (b) 1991 112 2008 116

1976 (b) 1992 109 2009 120

1977 95 1994 93 2010 116

1978 95 1995 94

1979 95 1996 91

a. Sources: Environmental reports in the References section.

b. Not reported.

Table 4-16. Ambient radiation by area (mrem/yr).

a,b

Area

Year 1 2 3 4 5 6 7 8 9 10 11 12 13 14

1967 318 194 205

1968 285 183 175 190

1969

1970

1971

1972

1973

1974

1975

1976

1977

1978 130 167 200 138 110 110 120 150 320 140 150

1979 119 147 190 140 133 104 125 115 140 168 155 191

1980 123 156 199 155 133 106 130 135 145 192 165 196

1981 120 155 218 150 166 113 125 150 180 160 187

1982 126 149 188 145 141 108 125 140 209 140 189

1983 115 141 181 143 144 103 95 135 201 140 174

1984 97 149 162 116 111 86 327 125 102 166 112 155

1985 141 78 101 107 347 94 114 225 180

1986 93 126 150 92 92 77 318 94 96 120 107 106

1987 146 189 184 164 191 126 120 149 181 133 168

1988 155 209 207 164 177 134 150 179 203 158 194

1989 140 217 205 132 141 106 126 151 179 153 139
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 33 of 88

Area

Year 1 2 3 4 5 6 7 8 9 10 11 12 13 14

1990 134 167 187 142 142 109 122 148 170 132 140

1991 128 168 194 146 165 116 132 124 132 150 130 190

1992 123 167 173 141 150 104 145 128 136 157 158 170

1993 147 178 200 158 198 134 165 153 156 172 184 201

1994 62 62 89 84 70 102 97

1995 102 136 150 121 212 87 176 102 104 179 124 130

1996 102 131 151 110 225 105 164 98 99 176 124 135

1997 131 86 88 81

1998 106 132 137 110 134 100 201 127 101 168 121 148

1999 108 133 135 122 110 105 191 129 99 160 122 152

2000 127 155 161 129 134 122 217 147 190 187 144 176

2001 121 299 151 120 136 107 214 143 115 180 140 163

2002 113 272 168 341 122 98 174 135 106 167 131 162

2003 180 260 257 334 145 104 159 209 189 188 133 169

2004 162 251 239 323 148 106 157 204 190 185 130 166

2005 159 246 236 320 148 108 159 205 191 182 132 166

2006 154 243 228 309 144 108 156 202 186 182 133 167

2007 155 242 228 306 147 112 158 199 190 184 138 171

2008 148 227 219 298 138 107 150 189 179 174 129 159

2009 151 229 218 283 140 109 152 191 180 175 134 164

2010 146 216 250 269 144 98 149 184 173 170 132 156

Area

Year 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 Average Maximum

1967 142 175 207 318

1968 186 131 172 189 285

1969 320c

1970 320c

1971 320c

1972 320c

1973 320c

1974 320c

1975 320c

1976 320c

1977 320c

1978 129 173 185 150 162 140 120 155 320

1979 121 150 145 155 160 115 117 152 130 120 125 140 191

1980 134 165 163 185 207 70 98 163 135 130 205 152 207

1981 138 155 168 174 163 75 126 175 150 140 152 218

1982 134 150 173 176 200 75 95 150 135 135 140 146 209

1983 124 135 151 158 206 125 163 149 140 185 148 206

1984 108 129 124 132 157 58 154 129 173 137 327

1985 115 123 131 127 103 153 143 347

1986 81 124 134 160 55 71 106 89 135 115 318

1987 135 151 174 246 204 68 96 129 139 155 246

1988 155 172 212 223 211 84 123 163 172 223

1989 162 160 168 178 189 78 116 146 118 150 217

1990 124 164 174 178 192 83 177 139 146 148 192

1991 122 165 175 173 189 81 79 143 154 110 155 144 194

1992 134 152 168 189 185 79 75 132 143 179 145 189

1993 145 177 195 192 213 91 96 158 169 195 167 213

1994 143 141 131 58 61 115 121 140 98 143

1995 109 129 137 149 153 58 70 117 121 127 212

1996 103 137 144 153 150 55 56 114 124 165 128 225
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 34 of 88

Area

Year 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 Average Maximum

1997 102 150 153 83 63 127 135 170 114 170

1998 99 135 151 156 76 58 113 123 165 127 201

1999 108 137 153 151 75 57 111 126 170 126 191

2000 117 160 185 182 92 81 139 158 205 153 217

2001 112 167 159 131 90 74 135 145 193 147 299

2002 103 144 172 157 81 68 122 128 146 341

2003 116 133 146 165 334 65 127 113 176 334

2004 111 137 145 158 309 93 123 112 172 323

2005 114 144 146 164 305 116 125 115 174 320

2006 117 143 149 161 290 76 125 115 169 309

2007 116 145 151 165 287 98 128 117 172 306

2008 111 137 144 161 272 68 124 114 162 298

2009 115 133 148 160 255 63 117 116 162 293

2010 111 140 142 157 247 68 116 114 152 269

a. Sources: Environmental Reports list in the References section.

b. Blank cells indicate data not available.

c. Ambient radiation not provided for these years; assumed equal to value for 1978.

Some covered employees remained on site continuously for weeks at a time. However, because 

most of the nonworking hours were spent indoors where elevated ambient radiation would be reduced 

by shielding from the building and because of the conservative assumptions used to estimate the 

values in Table 4-16, adjustment of the tabular data is not required to ensure ambient exposures are 

not underestimated for these individuals.

Although not explicitly stated in the NTS annual environmental reports, the overall uncertainty of 

measurements from modern TLDs (e.g., Panasonic UD-814) has been determined to be less than 

20% (ORAUT 2010a). However, for earlier measurement methods, the overall uncertainty could have 

been greater. Therefore, for dose reconstructions that are favorable to claimants, the assumption is 

made that the 95th-percentile values of the expected values (50th percentile) in Tables 4-15 and 4-16 

are enveloped by a factor of ±2 of the overall error of 50% (i.e., ±100%). Equation 4-1 (Section 

4.2.1.2.1) would then estimate the GSD for the expected values at 1.52.

Assignment of dosimeters to all persons at NTS began in April 1957 and ended in December 1992 

(ORAUT 2010a). During this time, readings from control badges at Mercury were used to subtract 

ambient background dose from the personnel dosimeters (ORAUT 2010a). For this reason, from April 

1957 to December 1992, ambient background external radiation dose should not be included in dose 

reconstructions. Because no documentation is available prior to 1967 (see Tables 4-15 and 4-16), 

ambient external doses should not be assigned under any circumstances from 1951 through 1956. 

After 1992, ambient background external dose should be assigned to all unmonitored workers in 

accordance with the guidelines for Tables 4-15 and 4-16.

4.3.2 Releases of Noble Gases

Ground seepage can increase when changes in ambient pressure pump small amounts of noble 

gases (primarily 85Kr and 133Xe) up through the overburden and into the atmosphere from the cavity 

created by a nuclear test. This process, sometimes referred to as atmospheric pumping, creates a 

diffuse source of radiological effluents.

In 1982, REECo assumed responsibility for six noble gas sampling stations EPA previously ran and 

replaced them with new samplers. These sampling units were housed in a metal toolbox with three 

metal air bottles attached with quick disconnect hoses. A vacuum was maintained on the first bottle, 

which caused a steady flow of air to be collected in the other two bottles. The flow rate was 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 35 of 88

approximately 0.5 cm3

/min. The two collection bottles were exchanged weekly and yielded a sample 

volume of about 3 × 105

 cm3 (Scoggins 1983).

The noble gases were separated and collected from the atmospheric sample by a series of cryogenic

gas chromatographic techniques. Water and carbon dioxide were removed at room temperature and 

krypton and xenon were collected on charcoal at liquid nitrogen temperatures. These gases were 

transferred to a molecular sieve where they were separated from remaining gases and each other. 

The krypton and xenon were transferred to separate scintillation vials and counted on a liquid 

scintillation counter. The lower limits of detection for krypton and xenon are 4 × 10-6 and 10 × 10-6

pCi/m3

, respectively (Scoggins 1983).

Krypton-85

The original six samplers were in Areas 1, 5, 12, 15, 23, and 25. These samplers showed no 

statistical difference in atmospheric concentrations of 85Kr during 1982 and 1983. The average 

concentration of all stations for these years was 24.6 pCi/m3 (0.91 Bq/m3

), which was established as 

NTS background. This background measurement shows good agreement with the global background 

of 27 pCi/m3 (1.0 Bq/m3

) (WMO 2002).

In 1983, REECo measured a statistically significant elevation in 85Kr concentrations in Area 20 of 

4.0 pCi/m3 (0.15 Bq/m3

) above background, which was subsequently related to seeps from nuclear 

cavities on Pahute Mesa (Scoggins 1984). These elevated concentrations continued until after the 

cessation of nuclear testing in 1992. Table 4-17 lists the average background and the Area 20 

elevated concentrations from 1983 through 1992.

Table 4-17. 85Kr average background and the Area 20 net concentrations and annual organ doses 

from 1983 through 1992.

a

Year

NTS 

background 

(pCi/m3)

Area 20 

camp 

(pCi/m3)

Net 

elevated 

(pCi/m3)

Sourceb elevated 

concentration 

(pCi/m3)

Sourceb elevated 

concentration 

(Bq/m3)

Other organc

dosed at sourceb

(mrem/yr)

Skin dosee

at sourceb

(mrem/yr)

1983 27 46 19 502 18.6 0.0029 0.18

1984 25 29 4 106 3.9 0.0006 0.04

1985 28 39 11 290 10.8 0.0017 0.10

1986 30 58 28 739 27.4 0.0043 0.26

1987 26 39 13 343 12.7 0.0020 0.12

1988 25 29 4 106 3.9 0.0006 0.04

1989 23 27 4 106 3.9 0.0006 0.04

1990 29 37 8 211 7.8 0.0012 0.07

1991 24 32 8 211 7.8 0.0012 0.07

1992 26 30 4 106 3.9 0.0006 0.04

1993 27 28 1 26 1.0 0.0002 0.01

a. Sources: Environmental reports in the References section.

b. Source receptor location assumed to be 500 m from U-20a emplacement hole.

c. Includes gonads, breast, lung, red marrow, bone surfaces, thyroid, remainder, and effective whole body.

d. Assumes 2,000 hr/yr exposure and dose conversion factor of 2.2 × 10-16 Sv/Bq/s/m3 (Eckerman and Ryman 1993).

e. Assumes 2,000 hr/yr exposure and dose conversion factor of 1.32 × 10-14 Sv/Bq/s/m3 (Eckerman and Ryman 1993).

The source of the elevated 85Kr concentrations was attributed to the emplacement hole nearest the 

Area 20 sampler (U-20a), which is about 3,660 m (12,000 ft) to the south. To determine if these 

elevated concentrations could be of any consequence for unmonitored workers, upper bound 

concentrations of 85Kr within 500 m of U-20a were estimated using Area 20 meteorological data 

station files and the CAP88-PC (ORAUT 2003) computer program. The program was used to develop 

atmospheric dispersion factors (X/Q) for various distances from the source of emissions. The ratio of 

the X/Q value at 500 m to that at 3,660 m was determined to be about 27, which means that the 85Kr 

concentration 500 m from the source could, on average, be 27 times greater than that measured at 

the sampler location. The potential dose resulting from submersion in the plume of this noble gas is 

entirely external because the gas is not readily assimilated by the body. Therefore, these source 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 36 of 88

concentrations were converted to potential organ doses using submersion dose conversion factors 

from Federal Guidance Report No. 12 (Eckerman and Ryman 1993). 

As Table 4-17 indicates, the upper bound annual organ doses for all years are less than 1 mrem. For 

reconstructing environmental doses for unmonitored employees, these doses are inconsequential for 

determination of POC and, therefore, should be ignored. In addition, if a monitored employee was 

exposed to the calculated upper bound 85Kr concentrations or significantly higher concentrations 

during work activities, these exposures would have been monitored by personal dosimeters and 

would, therefore, already be in the individual dosimetry record.

Uncertainty estimates for a number of 85Kr measurements indicate, generally, a standard deviation of 

less than 20% of the measured value. Even at the 2-sigma level of ±40 percent, the annual organ 

doses in Table 4-17 are less than 1 mrem and, therefore, inconsequential for employee dose 

reconstruction.

Xenon-133

Although the vast majority of the 133Xe measurements were below the lower limit of detection (i.e., 

10 × 10-6 pCi/m3

), elevated atmospheric concentrations of 133Xe were occasionally measured in 

Area 20. As with 85Kr, these elevated concentrations were attributed to seepage from underground 

test cavities at the Pahute and Rainier Mesa tests (Fauver 1985). However, 133Xe has a short half-life 

(5.25 days versus 10.7 years for 85Kr). Therefore, these elevated concentrations were short-lived. In 

addition, these elevated 133Xe concentrations were always less than the derived air concentrations 

(typically less than 3%) and would result in annual organ doses of less than 1 mrem. Therefore, 

potential doses to unmonitored employees from elevated concentrations of 133Xe are inconsequential 

for dose reconstruction.

4.4 INTERNAL INHALATION AND INGESTION DOSE TO UNDERGROUND WORKERS 

In the early 1980s, the Environmental Sciences Department of REECo recognized that the buildup of 

radon and radon daughter concentrations (RDCs) could pose a potential health problem in tunnels on 

Rainer Mesa and at other NTS locations. In 1984, to determine the concentrations of the RDCs and 

the effect of environmental conditions on the buildup of these concentrations, REECo conducted 

radon measurement surveys in G-, T-, and N-Tunnels (Fauvor 1987). This section discusses the 

results of these surveys and those in 1991 and 1992 (Lyons 1992a,b).

4.4.1 Underground Activities

Area 12, which is in the Nuclear or High Explosive Test Zone, occupies 104 km2 (40 mi2) at the 

northern boundary of NTS, known as Rainier Mesa. No atmospheric nuclear tests have occurred at 

this location; however, Area 12 was the site of the U.S. Atomic Energy Commission’s first fully 

contained underground nuclear detonation, the RAINIER test, on September 19, 1957, in a horizontal 

tunnel about 487 m (1,600 ft) into the mesa and 274 m (900 ft) beneath the top of the mesa (ORAUT 

2008). In the past several decades, a number of tunnels have been mined into Rainier Mesa, in 

which most of the U.S. Department of Defense (DOD) horizontal line-of-sight exposure experiments 

have occurred. The N-, P-, and T-Tunnel complexes, in particular, were developed extensively during 

the 1970s and 1980s. The tunnel experiments usually involved complex construction of largediameter [up to 9 m (27 ft)], line-of-sight pipes and special closure mechanisms, blast and gas seal 

doors, stemming plugs, and the like. The G-Tunnel complex was originally established for nuclear 

testing but, since 1971, has been used only as an underground research facility (ORAUT 2008). 

In addition to its use for nuclear testing purposes, N-Tunnel was the location of a Nonproliferation 

Experiment involving 1.3 million kg (2.9 million lb) of conventional explosives on September 22, 1993. 

DOD operates a high-explosives research and development tunnel in Area 12. This reusable test bed 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 37 of 88

supports programs involving the detonation of conventional or prototype high explosives and 

munitions (ORAUT 2008). 

The U1a Complex in Area 1 is an underground laboratory of horizontal tunnels about 0.5 mi long at 

the base of a vertical shaft about 960 ft beneath the surface. The vertical shaft is equipped with a 

mechanical hoist for personnel and equipment access. Another vertical shaft about 1,000 ft away 

provides cross ventilation, instrumentation, utility access, and emergency egress. The shaft was 

excavated in the 1960s, and a nuclear test was conducted in a horizontal tunnel mined from its base 

in 1990 (ORAUT 2008).

4.4.2 Radon Measurements

Radon-222, with a radioactive half-life of 3.8 days, occurs in the 238U decay chain; 220Rn, with a 

radioactive half-life of 54.5 seconds, occurs in the 232Th chain. The decay daughters associated with 

222Rn include 218Po, 214Pb, 214Bi, and 214Po, and those associated with 220Rn include 216Po, 212Pb, 212Bi, 

and 208Tl. Because radon is a noble gas and thus chemically inert, it migrates easily from the tunnel 

rock and soil, which contains naturally occurring trace quantities of uranium and thorium.

Radon progeny measurements in N- and T-tunnels were made in drifts mined in a rock formation 

known as Tunnel Bed Non-Welded Ash Fall Tuff. This rock is unconsolidated and only slightly 

fractured. The drift that was sampled in the inclined G-tunnel was mined in Grouse Canyon Welded 

Ash Fall Tuff, which is extremely fractured. Factors that affect concentrations of radon and radon 

progeny in air are ventilation rates, barometric pressure, relative humidity, temperature inversions, the 

degree of fracturing in the rock, and the amount of smoke and dust in the air. Factors such as 

barometric pressure and fractures affect the rate at which radon emanates from the rock, while others 

such as dust and ventilation rates affect the accumulation rate of progeny in the air. Concentrations 

fluctuate with changing seasons, weather conditions, and activities in the area being monitored 

(Fauver 1985).

The concentration of radon progeny (the major dose contributors) in air is measured in working levels

(WL). This is the common unit for expressing radon progeny exposure rates. The WL was developed 

for use in uranium mines but is now used for environmental exposures. Numerically, the WL is any 

combination of short-lived decay products in 1 L of air that results in the emission of 1.3 × 105 MeV of 

potential alpha energy. When radon is in complete equilibrium with its short-lived decay products, 

1 WL equals 100 pCi/L (i.e., 100 pCi/L each of 222Rn and short-lived decay products 218Po, 214Pb, 214Bi, 

and 214Po) (NCRP 1988, p. 17). For 220Rn and its decay products (216Po, 212Pb, 212Bi, and 208Tl), 1 WL 

is equal to 7.47 pCi/L. The advantage of the WL unit is that it enables comparison of different 

equilibrium levels and different concentrations of radon decay products. The degree of equilibrium is 

a critical factor for estimating inhalation exposure and is of equal importance to the radon 

concentration itself (NCRP 1988, p. 19). The WL unit considers this factor. However, because the 

dose conversion factors for 220Rn progeny (212Pb + 212Po) are one-third the values for the 222Rn 

progeny (ICRP 1981), the dose equivalent WL for 220Rn is only one-third that of a 222Rn WL.

The exposure of tunnel workers can be expressed in units of working-level months (WLM), which is 

an exposure rate of 1 WL for a working month of 170 hours (NCRP 1988, p. 17). For example, an 

exposure of 1 WLM would result from exposure to a concentration of 1 WL for 1 month or 0.5 WL for 

2 months. 

4.4.3 Underground Radon Concentrations

Measurements were taken in tunnels N, T, and G. The preliminary measurements indicated N- and 

T-Tunnel RDCs of about 0.01 WL under normal ventilation conditions. However, the data 

demonstrated that RDCs could rise to relatively high levels (i.e., 0.24 WL) when ventilation rates were 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 38 of 88

significantly lower (Fauvor 1987, Figure 2, p. 7). The radon daughter concentrations in G-Tunnel 

were an order of magnitude higher than those in N- and T-Tunnels. The average RDC in the rock 

mechanics drift (the worst-case location in G-Tunnel) was 0.13 WL (ranging from 0.07 to 0.23 WL). 

Elevated RDCs in the rock mechanics drift of G-Tunnel seemed to be from a lower ventilation rate in 

conjunction with the more highly fractured nature of the welded tuff in which the incline drift was 

mined. By increasing the ventilation rate, a 60% reduction in RDCs from an average of 0.13 WL to an 

average of 0.05 WL was achieved (Table 4-18). 

Table 4-18. Results of experiment to determine the effects of ventilation 

conditions on RDCs in the G-Tunnel inclined drift.

a

Location

Ventilation 

conditions

Radon daughter integrated 

sample average (WL)

Rock mechanics drift at 0+52 Alternating 0.19

Continuous 0.05

Average Alternating 0.13

Continuous 0.03

a. Source: Fauvor (1985, Table 3). 

Lyons (1992a,b) reported additional radon measurements from 1991 and 1992. Although the 

G-Tunnel complex remained inactive in 1992, radon samples were taken to document potential radon 

WL in a worst-case scenario of complete ventilation failure throughout the complex. The maximum 

WL for radon daughters was 1.4. No samples were taken in 1992 in T-Tunnel, which was inactive 

that year.

Tables 4-19 and 4-20 list the results of extensive radon sampling in N- and P-Tunnel complexes in 

1992 (from Lyons 1992a, Table 3, p. 8, and Lyons 1992b, Table 1, p. 8, respectively). The average 

concentrations in N- and P-Tunnels were 0.021 and 0.009 WL, respectively, and the maximum 

concentrations in N- and P-Tunnel were 0.038 and 0.017 WL, respectively.

4.4.4 Underground Worker Exposure to Radon

Although measurements were periodically performed in the tunnel complexes to ensure adequate 

worker protection, neither DOE nor its predecessor agencies attempted to quantify or record 

occupational exposures to radon and its daughters. Therefore, dose reconstructors should adjust the 

dose for lung, ET1, and ET2 cancers of any employee who was a miner or tunnel worker to account 

for radon exposure while working in the tunnel complexes. To quantify the exposure, Table 4-21 lists 

airborne 222Rn RDCs based on measurement results that are favorable to claimants. The values in 

Table 4-21 pertain only to 222Rn and its daughter products. The 220Rn WL concentrations and 

resultant exposures for N- and P-Tunnels must be converted to annual alpha organ dose before entry 

in IREP.

For G-Tunnel workers, annual exposures (assuming a full year of underground activity) are assumed 

to be 1.92 WLM before 1984 and 0.60 WLM during and after 1984. This is because REECo did not 

recognize until 1984 that significant reductions in RDCs could be effected by leaving all ventilation 

fans running overnight. Until 1984, the practice was to shut down alternating fans each night, which 

resulted in higher average RDCs.
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 39 of 88

Table 4-19. Radon daughter concentrations for N-Tunnel in 1991 and 

1992. 

N-Tunnel location

Rn-222 (WL) Rn-220 (WL)

Average Maximum Average Maximum

January–June 1992

Miner's lunchroom 0.005 0.007 0.01 0.016

Raytheon Alcove 0.003 0.004 0.005 0.008

Slow Alcove 0.005 0.007 0.008 0.009

24 Bypass Drift 0.005 0.006 0.015 0.015

24 LOS Drift at GZ 0.005 0.007 0.012 0.014

22 Bypass 0.006 0.007 0.013 0.015

July–December 1991

Miner's lunchroom 0.005 0.007 0.011 0.018

Raytheon Alcove 0.009 0.03 0.017 0.06

21 LOS at 2 + 50 0.034 0.059 0.029 0.046

15 Assembly Drift 0.006 0.009 0.015 0.021

Slow Alcove 0.004 0.008 0.01 0.025

23 Fast Alcove 0.009 0.014 0.018 0.03

24 Bypass Drift 0.005 0.007 0.012 0.041

24 LOS Drift 0.005 0.01 0.015 0.036

Average 0.008 0.013 0.014 0.025

Table 4-20. Radon daughter concentrations for P-Tunnel in 1991 and 

1992. 

P-Tunnel location

Rn-222 (WL) Rn-220 (WL)

Average Maximum Average Maximum

January–June 1992

01 Drift at Access Drift 0.001 0.001 0.002 0.005

01 Fast Alcove 0.001 0.001 0.004 0.006

02 Main Drift at 6 + 00 0.001 0.001 0.002 0.004

04 Reentry 0.001 0.002 0.004 0.005

04 LOS at 12 + 00 0.002 0.002 0.005 0.006

HPD Base Station 0.001 0.001 0.003 0.006

Miner's lunchroom 0.001 0.001 0.003 0.006

05 Cavity 0.001 0.001 0.003 0.004

July–December 1991

01 Drift at Access Drift 0.003 0.006 0.004 0.007

01 Fast Alcove 0.003 0.005 0.003 0.006

02 Main Drift at 6 + 00 0.01 0.032 0.01 0.046

04 LOS at VP X-Cut 0.008 0.013 0.012 0.015

04 LOS Drift at GZ 0.005 0.006 0.014 0.02

04 LOS Test Ch. 0.003 0.004 0.005 0.006

04 Bypass at RE#1 0.007 0.013 0.008 0.011

IHD Alcove 0.003 0.009 0.004 0.015

LLNL Alcove 0.003 0.006 0.007 0.02

05 Cavity 0.003 0.006 0.006 0.015

Average 0.003 0.006 0.006 0.011

Because no radon measurement data were readily available for the underground portions of the U1a 

Complex, the assumption, which is favorable to claimants, was made that the RDCs in U1a would be 

similar to those in G-Tunnel (which is mined in highly fractured tuff) [1]. In addition, a reasonable 

assumption was made that the U1a Complex is well ventilated, similar to G-Tunnel. Dose 

reconstructors should use these average annual exposure values for miners and tunnel workers 

without identified tunnel locations. 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 40 of 88

Table 4-21. Annual 222Rn exposures and uncertainties for internal dose reconstruction 

for miners and tunnel workers (Fauver 1985). 

Tunnel complex

RDC concentration 

(WL)

Annual exposure

(WLM)a

Uncertainty

GSDb

G (before 1984) 0.16 1.92 1.52

G (1984 and later) 0.05 0.60 1.52

N 0.013 0.16 1.52

P 0.006 0.07 1.52

T 0.01 0.12 1.52

Unidentifiedb (before 1984) 0.16 1.92 1.52

Unidentifiedb (1984 and later) 0.05 0.60 1.52

a. Based on 170 hours of exposure per month for 12 months.

b. Use these values if underground work location is not known.

For dose reconstruction, the GSD of the values in Table 4-21 can be estimated using Equation 4-1 

(Section 4.2.1.2.1). Under the assumption that the 50th-percentile expected annual exposures are 

those in Table 4-21 and that the 95th-percentile values are twice the values in Table 4-21, the GSD 

for all values in Table 4-21 is 1.52. 

Gravel Gerties

As discussed in Part 2, Section 2.2.5 of this site profile (ORAUT 2008), underground bunkers were 

constructed in Area 5 for testing containment capabilities involving accidental explosion of highexplosive material in the underground bunker. Containment tests were conducted in the Gravel 

Gerties in 1957 and again in 1982. These tests resulted in successful designs that were used at the 

Pantex Plant in Texas and in the Device Assembly Facility in Area 6. Five Gravel Gerties were 

constructed in Area 6 from 1988 to 1992. However, with the cessation of testing in 1992, they have 

not been activated for device assembly.

Because the Area 5 Gravel Gerties were constructed for testing and the Area 6 Gravel Gerties were 

never activated, it is not likely that workers spent significant time inside the structures. However, if it 

can be determined from the dosimetry records or telephone interviews that a claimant worked in the 

Gravel Gerties for a significant time, it should be assumed that they would have been exposed to 

elevated levels of radon, and this radon exposure should be evaluated and included in the dose 

reconstruction.

Although radon measurements are not available for inside the NTS Gravel Gerties, because of the 

similarities of construction, application of the radon concentrations measured in the Pantex Gravel 

Gerties is assumed to be appropriate. Therefore, for a 12-month period for which it can be 

determined that a claimant worked in the NTS Gravel Gerties, a radon exposure of 0.072 WLM should 

be assigned (ORAUT 2007a, Section 5.3.3). This exposure can be prorated as appropriate for 

employment periods of less than a full year. These exposures should be applied as lognormal with a 

GSD of 3 (ORAUT 2007a, Section 5.3.3).

Uncertainty Associated with Radon Measurements

Two methods used to measure the concentration of radon and its progeny were the grab sample 

technique and the integration technique. Preliminary measurements were made by grab sampling for 

a general estimate of the concentrations. Lucas cells collected and counted grab samples of radon 

gas. The cells were evacuated in the laboratory and opened at the sampling location for a single 

intake of air. The filled cells were counted at the EPA Las Vegas Laboratory within 24 hours to 

minimize decay. The number of measurements was small due to limited availability of Lucas cells. 

Grab samples of radon daughters were collected and counted according to a technique described by 

Rolle (ANSI 1973), which utilized a single count of an air sample collected on a filter. Samples were 

collected on a 2-cm Whatman fiberglass filter at a rate of about 5 L/min and counted on an EDA 

Instruments Incorporated Radon Detector (Model RD200). The Rolle method allowed a choice of 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 41 of 88

several different analysis regimes. The regime chosen for tunnel measurements was a 5-minute 

sampling time, a 6- minute decay time, and a 5- minute counting time. Using appropriate correction 

factors, the RDC in WL can be evaluated 11 minutes after sample collection. The values obtained 

using the Rolle method were periodically checked using the Kusnetz (Fauvor 1987, p. 3) method, 

which also used a single count of an air filter but allowed for a 40-to-90-minute decay time. These 

measurement techniques have good accuracy and a relative standard deviation of less than 15% 

(Fauvor 1987, p. 3). 

Grab sampling is quick and convenient. However, as mentioned above, radon concentrations can 

fluctuate widely with time and location. To account for this variability, integrating monitoring 

instruments were used. A passive environmental radon monitor (PERM) was used to measure radon 

in picocuries per liter. A radon progeny integrating sampling unit (RPISU) measured the RDC in WL.

The PERM is an integrating radon monitor that employs electrostatic collection of radon daughter ions 

and uses a TLD as the radiation dose integration element. The lower limit of detection at the 95% 

confidence level is about 0.3 pCi/L for a 1-week sample with a relative standard deviation of 20% 

(Fauvor 1987, p. 4). The RPISU is a low-volume air sampler that draws air through a 0.65-μm-poresize Millipore membrane filter. A Teflon disc TLD containing dysprosium-activated calcium fluoride 

(Harshaw TLD200) is positioned close to the collection surface of the filter and serves as the dose 

integrating element. The lower limit of detection at the 95% confidence level is about 0.0001 WL for a 

1-week sampling period with a relative standard deviation of 10% (Fauvor 1987, p. 4). The RPISU 

technique is recommended for the measurement of RDCs by the DOE Technical Measurements 

Center, Division of Remedial Action Projects (Fauvor 1987, p. 4).

4.4.5 Underground Worker Exposure to Other Radionuclides

In addition to exposure to elevated levels of radon, underground workers could have been exposed to 

other sources of airborne contamination. Because the breathing air in the underground environment 

was drawn from the outside environment, underground workers would have been exposed to the 

same inhalation intakes as aboveground workers (see Sections 4.2.1.2.2 through 4.2.1.2.6). 

4.5 INSTRUCTION TO DOSE RECONSTRUCTORS FOR ASSIGNMENT OF 

ENVIRONMENTAL INTAKES 

4.5.1 Assigning External Ambient Dose

Starting in April 1957 and ending in 1992, all personnel entering NTS were monitored for external 

radiation exposure (ORAUT 2010a). Because unexposed control badges were processed with the 

personnel dosimeters and the readings from the control badges were subtracted from the dosimeter 

to obtain a net reading for determining exposure, ambient external dose should not be included in 

dose reconstruction before 1993. Before the institution of universal badging in April 1957 for 

employees who were not monitored, coworker doses should be assigned based on the values in 

ORAUT (2010a, Table 6-11) and no additional ambient dose should be assigned through 1992. 

Based on a review of the dosimetry literature of the site, no change in the handling of dosimeters 

occurred when universal badging was introduced; therefore, ambient dose is appropriately included in 

the coworker doses beginning in 1957 through 1992. After 1992, ambient external dose should be 

added to dose reconstruction for unmonitored workers in accordance with the guidelines for assigning 

dose from Tables 4-15 and 4-16.

External ambient doses for each year are available in Tables 4-15 and 4-16 and can be used for 

workers with continuous employment but nonexistent or incomplete monitoring records. For example, 

EG&G employees were likely to be stationed in Las Vegas and received dosimeters when they 

traveled to NTS. If the employee’s location cannot be clearly identified, it might be appropriate to 
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 42 of 88

assign ambient dose for intervals with no dosimeter exchanges. A second example includes REECo 

employees who were processed as new employees through NTS (Mercury) and assigned to work at 

Tonopah Test Range (TTR). These employees have at least one NTS dosimeter. Often no additional 

records are provided for these employees by Sandia National Laboratories, the responsible 

organization for dosimetry at TTR. For REECo employees assigned at TTR, ambient dose (ORAUT 

2007b) can be assigned as appropriate. 

4.5.2 Assigning Internal Ambient Dose

With the exception of cases that can be worked using the bounding assumptions in ORAUT-OTIB0018 (ORAUT 2005), environmental inhalation and ingestion intakes listed in Tables 4-7 and 4-11, 

respectively, shall be applied starting in 1964. 

To correct for exposure to short-lived fission and activation, the dose reconstructor should add annual 

doses that were greater than 0.001 rem from Tables 4-9 and 4-14 as 30-to-250-keV photons with a 

constant distribution. These doses shall be entered into the IREP analysis as constants because they 

can be considered bounding overestimates of the actual doses.

4.6 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

[1] Rollins, Eugene M. Oak Ridge Associated Universities Team. Division Manager. October 

2007. The highest radon concentrations were measured in the G-Tunnel and occurred in 

fractured tuff. Therefore, with the lack of measured data in the U1a Complex, an assumption 

favorable to claimants was made that the radon concentrations in the U1a Complex were the 

same as the highest measured concentrations in the G-Tunnel complex.
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 43 of 88

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Townsend, Y. E., and R. F. Grossman, editors, 2000, Nevada Test Site Annual Site Environmental 

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Wills, C. A., 2004, Nevada Test Site Environmental Report 2003, DOE/NV/11718-971, A. L. McCurdy, 

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Wills, C. A., 2006, Nevada Test Site Environmental Report 2005, DOE/NV/11718-1214, 

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Wruble, D. T., and E. M. McDowell, editors, 1990a, Nevada Test Site Annual Site Environmental 

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Hicks, H. G., 1981e, Calculation of the Concentration of any Radionuclide Deposited on the Ground 

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McArthur, R. D., and S. W. Mead, 1987, Nevada Test Site Radionuclide Inventory and Distribution 

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for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

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GLOSSARY

absorption

In external dosimetry, process in which radiation energy is imparted to material. In internal 

dosimetry, movement of material to blood regardless of mechanism.

absorption type

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively in the 

respiratory tract (slow solubilization). Also called solubility type.

accuracy

The characteristics of an analysis or determination that ensures that both the bias and 

precision of the resultant quantity will remain within the specified limits.

alpha particle (α)

See alpha radiation. 

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

ambient gamma radiation

Penetrating gamma radiation in the outside environment. Includes gamma radiation from 

natural cosmic and terrestrial sources as well as synthetic sources.

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

beta particle (β)

See beta radiation. 

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

breathing rate

Amount of air a person breathes in a specified time. In relation to health physics for workers, 

rates typically vary from light (1.2 cubic meters per hour) to heavy (1.7 cubic meters per hour) 

as defined by the International Commission on Radiological Protection. 

buildup

Increase in flux or dose due to scattering in the medium.
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confidence level

The interval about an estimate of a stated quantity within which the value of the quantity is 

expected to be with a specified probability. See uncertainty. 

contamination

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, or grays.

dose conversion factor

Multiplier for conversion of potential dose to the personal dose equivalent to the organ of 

interest (e.g., liver or colon).

dose equivalent (H) 

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose. 

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See film dosimeter and thermoluminescent dosimeter. 

dosimeter holder

Plastic holder for a dosimeter card, which typically includes one or more metallic filters that 

modify the response of the phosphor to radiation.

dosimetry

Measurement and calculation of internal and external radiation doses.

element

One of the known chemical substances in which the atoms have the same number of protons. 

Elements cannot be broken down further without changing their chemical properties.

environmental occupational dose

Dose received from radiation site-related activities (i.e., above normal background levels) 

while on a site, which is often recorded by monitoring stations in specific areas or along the 

boundaries of facilities.

error

Difference between the correct, true, or conventionally accepted value and the measured or 

estimated value. Sometimes used to mean estimated uncertainty. See accuracy and 

uncertainty. 

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.
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exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

external dose

Dose received from radiation emitted by sources outside the body.

favorable to claimants

In relation to dose reconstruction for probability of causation analysis, having the property of 

ensuring that there is no underestimation of potential dose, which often means the assumption 

of a value that indicates a higher dose than is likely to have actually occurred in the absence of 

more accurate information. See probability of causation. 

film

In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. 

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

filter

Material used in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

gamma ray, particle, or photon (γ)

See gamma radiation. 

geometric standard deviation (GSD)

In probability theory and statistics, the geometric standard deviation describes the spread of a 

set of numbers whose preferred average is the geometric mean. 

Gravel Gertie

At NTS, a round room with 2-foot-thick concrete walls and a staging area with two staging 

bays connected to the round room. Tests were conducted in the Gravel Gertie with high 

explosives and uranium devices to measure fallout.

high explosive

Chemical compound or mechanical mixture that, when subjected to heat, impact, friction, 

shock, or other initiation stimulus, undergoes a rapid chemical change resulting in large 

volumes of highly heated gases that exert pressure in the surrounding medium. See 

insensitive high explosive. 

holder

See dosimeter holder. 
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ingestion

Process of taking a substance into the body through the mouth.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties. See element. 

kiloelectron-volt (keV)

Unit of particle energy equal to 1,000 (1 × 103

) electron-volts.

megaelectron-volt (MeV)

Unit of particle energy equal to 1 million (1 × 106

) electron-volts.

monitor

Device that detects and usually records the presence of radioactivity or radioactive 

substances.

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron. 

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.
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photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Program Act, the percent likelihood, at the 99th percentile, that a worker 

incurred a particular cancer from occupational exposure to radiation.

progeny

Nuclides that result from decay of other nuclides. Also called decay products and formerly 

called daughter products.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation. 

radioactive

Of, caused by, or exhibiting radioactivity. 

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. See radionuclide. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

roentgen

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts).

scattering

Change in direction of radiation by refraction or reflection, often accompanied by a decrease in 

radiation due to absorption by the refracting or reflecting material.

scintillation counter or detector

Combination of phosphor, a photomultiplier tube, and the associated electronic circuits for 

measuring the light emissions produced in the phosphor by ionizing radiation.

seep

At NTS, uncontrolled slow release of radioactive material with little or no energy. Seeps are 

not visible and can be detected only by measuring for radiation.
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shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

sievert (Sv)

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose.

thermoluminescent dosimeter chip

Small block or crystal of lithium fluoride in a thermoluminescent dosimeter. A TLD-600 

dosimeter contains a chip made from more than 95% 6

Li for neutron radiation detection, and a 

TLD-700 dosimeter contains a chip made from more than 99.9% 7

Li for photon and beta 

radiation detection. Also called crystals.

transuranic (TRU) elements

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium. 

uncertainty

Standard deviation of the mean of a set of measurements. The standard error reduces to the 

standard deviation of the measurement when there is only one determination. See accuracy, 

confidence level, and error. Also called standard error.

unmonitored dose

Potential unrecorded dose that could have resulted because a worker was not monitored.

venting

At NTS, prompt, massive, and uncontrolled releases of radioactive material characterized as 

active releases under pressure, such as when radioactive material is driven out of the ground 

by steam or gas.

weighting factor

The ratio of the stochastic risk arising from tissue T to the total risk when the whole body is 

irradiated uniformly.

whole-body (WB) dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose. 

working level (WL)

Unit of concentration in air of the short-lived decay products of 222Rn (218Po, 214Pb, 214Bi, and 

214Po) and 220Rn (216Po, 212Pb, 212Bi, 212Po) defined as any combination of the short-lived 

radioactive progeny of radon or thoron in 1 liter of air, without regard to the degree of 

equilibrium, that results in the ultimate emission of 130,000 MeV of alpha energy; 1 WL equals 

2.083 × 10-5 joules per cubic meter.
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 57 of 88

working level month (WLM)

Unit of exposure to radon progeny defined as exposure for 1 working month (170 working 

hours) to a potential alpha energy concentration from of 1 WL; 1 WLM equals 1 WL times 

170 hours, which is 0.00354 joule-hours per cubic meter. See working level. 

X-ray

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 58 of 88

TABLE OF CONTENTS 

SECTION TITLE PAGE

A.1 Introduction ............................................................................................................................ 59

A.2 Air Sampling Data and Intake Estimates ................................................................................ 60

A.3 Soil Contamination and Decay Corrections ............................................................................ 63

A.4 Scaling Factors for Inhalation Intake Estimates...................................................................... 65

A.5 Correction for Resuspension for Early Times After Atmospheric Tests................................... 65

A.6 Corrections for Inhalation Dose From Short-lived Fission and Activation Products................. 68

A.7 Ingestion Pathway.................................................................................................................. 76

A.8 Instruction to Dose Reconstructors for Assignment of Environmental Intakes ........................ 84

LIST OF TABLES

TABLE TITLE PAGE

A-1 Atmospheric concentrations of 239Pu for sampled areas......................................................... 61

A-2 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 62

A-3 Inventory of contaminated soil................................................................................................ 64

A-4 Radionuclide areal soil deposition.......................................................................................... 64

A-5 Radionuclide areal soil deposition decay, corrected to 1963 .................................................. 65

A-6 Abundance of radionuclides in NTS soils relative to 239Pu decay, corrected to 1963 .............. 66

A-7 Scaled inhalation intakes corrected for early resuspension in 1963........................................ 67

A-8 Far-field refractory enrichment ratios...................................................................................... 69

A-9 Organ-specific inhalation dose fission and activation product correction factors..................... 72

A-10 Inhalation dose from short-lived fission and activation products ............................................. 76

A-11 Radionuclide soil concentration by area................................................................................. 76

A-12 Area-specific and maximum annual ingestion rates................................................................ 77

A-13 Annual organ doses from ingestion of 238,238Pu and 241Am...................................................... 78

A-14 Organ-specific ingestion fission and activation correction factors........................................... 83

A-15 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/yr ............................... 83

A-16 Organ-specific doses from ingestion of fission and activation products .................................. 84

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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LIST OF FIGURES

FIGURE TITLE PAGE

A-1 Resuspension factor as a function of time after initial deposition............................................ 67

A-2 SMALL BOY – 90Sr fraction of total dose lungs ...................................................................... 70

A-3 LITTLE FELLER I – 90Sr fraction of total dose to lungs........................................................... 71

A-4 TURK – 90Sr fraction of total dose to lungs............................................................................. 71

A-5 Strontium-90 fraction of total dose.......................................................................................... 72

A-6 Strontium-90 fraction of total dose.......................................................................................... 73

A-7 Strontium-90 fraction of total dose.......................................................................................... 73

A-8 Strontium-90 fraction of total dose.......................................................................................... 74

A-9 Strontium-90 fraction of total dose to the liver ........................................................................ 74

A-10 Strontium-90 fraction of total dose to the red marrow............................................................. 75

A-11 Strontium-90 fraction of total dose to bone surface ................................................................ 75

A-12 Strontium-90 fraction of total ingestion dose .......................................................................... 80

A-13 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 80

A-14 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 81

A-15 Strontium-90 fraction of total ingestion dose .......................................................................... 81

A-16 Strontium-90 fraction of ingestion dose to the upper large intestine ....................................... 82

A-17 Strontium-90 fraction of total ingestion dose .......................................................................... 82

A.1 Introduction

With the cessation of atmospheric testing at NTS in 1962, the greatest potential for environmental 

intakes of radioactive material results from the inhalation of radioactive particles that were 

resuspended from NTS soils into the atmosphere and from ingestion of soils contaminated by 

atmospheric nuclear weapons tests, reactor tests, and safety tests. The potential inhalation intakes 

can be estimated from air sampling data in the NTS annual environmental reports (see References 

section) coupled with extensive soil contamination data from between 1983 and 1991 (McArthur and 

Kordas 1983, 1985; McArthur and Mead 1987, 1988, 1989; McArthur 1991). Because the air 

monitoring data were limited to gross alpha and beta measurements, tritium, and isotopes of 

plutonium (e.g., 238Pu, 239Pu, and 240Pu), inhalation intakes of other relatively long-lived radionuclides 

that have been identified in NTS soils (e.g., 241Am, 60Co, 137Cs, 90Sr, 152Eu, 154Eu, and 155Eu) are 

scaled to those of plutonium based on their relative abundance in NTS soils. Ingestion intakes can be 

estimated by assuming consumption of contaminated NTS soils. To ensure that inhalation and 

ingestion intakes are not underestimated, the relative abundances of long-lived radionuclides in NTS 

soils that were determined from the 1991 soil contamination data (McArthur 1991) were decaycorrected to 1963. In addition, to ensure that intakes and resultant doses are not underestimated, 

correction factors are developed to account for potential exposures to short-lived fission and activation 

products based on test-specific data from Hicks (1981a,b,c,d). In addition, a correction factor is 

developed for inhalation intakes which accounts for the phenomenon of early resuspension 

(Anspaugh et al. 2002).

This approach has been developed to address the unique character of NTS, which is a large outdoor 

testing facility where potential exposure to radioactive materials was primarily based on residual 

contamination from atmospheric testing. For most employees of the primary contractor, occupational 

ATTACHMENT A

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BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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internal doses can be equated to the doses they potentially received from the resuspension of 

residual radioactivity from the atmospheric testing (i.e., environmental internal dose). For workers 

who were primarily indoors (cafeterias, administrative facilities, and maintenance shops) this dose is 

likely to represent an overestimate of internal exposure. The environmental internal dose is the 

occupational internal dose, regardless of the recorded external dose. If an internal exposure was 

suspected, bioassay was performed. Managing radioactive material in the form of devices was 

episodic and limited to a few workers (e.g., radiation safety and industrial hygiene personnel, miners, 

and experimenters). These workers are identified on the rosters that were published before the event, 

and these workers are likely to have bioassay results in the DOE records.

The following paragraphs discuss:

• Air sampling data from the NTS annual environmental reports from 1971 through 2001 and the 

calculated annual inhalation 239Pu intakes. 

• NTS soil contamination data that were gathered and reduced in the late 1980s (McArthur 

1991); soil contamination data that are decay-corrected to 1963. 

• Methods used to develop inhalation intake scaling factors for radionuclides identified in NTS 

soils in comparison with intakes of plutonium that were calculated from air sampling data. 

• The method used to account for higher air concentrations due to resuspension for 1963 (the 

first year after suspension of atmospheric testing). 

• The methods used to develop the correction factors that were used to account for organ dose 

from inhalation of short-lived radionuclides that no longer persist in the NTS environment in 

measurable quantities. 

• The method used to calculate ingestion intakes from oral consumption of contaminated NTS 

soils; the method used to account for ingestion dose from short-lived fission and activation 

products. 

• Step-by-step instructions to the dose reconstructor for application of ambient environmental 

inhalation and ingestion intakes. 

Attachment B provides a discussion that includes the results of 30 years of inhalation and ingestion 

intakes. 

A.2 Air Sampling Data and Intake Estimates 

Routine isotopic atmospheric measurements of plutonium at NTS began in 1971 with samplers in 15 

locations across the site. Six additional sampling stations were added in 1978. Equipment at fixed 

locations continuously sampled the ambient air to monitor for radioactive materials. These locations 

were chosen to provide representative samples from the populated areas on the site and to monitor 

resuspension of low-fired plutonium spread by safety experiments before 1960 in Areas 2, 3, 4, 7, 9, 

and 10. Access, worker population, geographical coverage, presence of radioactivity, and availability 

of electric power were considerations in the site selection for air samplers (Black and Townsend 

1997).

ATTACHMENT A

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BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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In 1988, efforts to monitor radioactive air emissions at NTS were increased as a result of the 

requirements of DOE Order 5400.1 (DOE 1990). Known and potential effluent sources throughout 

NTS were assessed for their potential to contribute to public dose (Black and Townsend 1997).

The 239Pu concentrations in Table A-1 for 1989 through 2001 represent the average of the maximum 

concentrations for a given area in a given year. If maximum values were not provided (i.e., 1971 

through 1988), the average of the average concentrations was reported. In addition to the annual 

area concentrations, Table A-1 lists NTS site average and maximum concentrations, which represent 

the arithmetic averages of the concentration of all the areas and the maximum of the area maximum 

or area averages of all the areas, respectively. Potential intakes associated with these concentrations 

can be calculated under the assumption that an unmonitored worker was occupationally exposed for 

2,000 hr/yr with a breathing rate of 2,400 m3

/yr. Table A-2 lists these calculated intakes.

Table A-1. Atmospheric concentrations of 239Pu for sampled areas (pCi/m3

).a

Year

Area

1 2 3 4 5 6 7 9 10 11 12

1971 3.7E-04 6.6E-04 1.7E-04 1.2E-04 1.7E-04 7.2E-04 2.1E-04 9.1E-05 1.6E-04

1972 1.5E-04 2.2E-04 3.7E-04 1.4E-04 1.3E-03 4.3E-03 2.9E-04 2.4E-04 7.9E-04

1973 8.7E-05 2.2E-04 2.1E-04 8.7E-05 1.2E-04 8.6E-04 4.4E-05 2.4E-04 4.3E-05

1974 8.0E-05 6.8E-05 1.3E-04 5.0E-05 7.9E-05 2.1E-04 5.7E-05 8.0E-05 5.8E-05

1975 2.9E-05 3.8E-05 1.5E-04 4.3E-05 6.3E-05 1.7E-04 5.7E-05 4.9E-05 3.3E-05

1976 2.9E-04 1.7E-04 1.0E-03 1.5E-04 4.4E-04 3.2E-03 2.1E-04 2.4E-04 2.8E-04

1977 2.7E-05 6.5E-05 2.6E-04 3.4E-05 5.5E-05 2.5E-04 4.6E-05 1.2E-04 2.1E-05

1978 4.7E-05 1.3E-04 2.8E-04 6.1E-05 9.9E-05 5.4E-04 2.9E-04 1.1E-04 6.2E-05

1979 7.4E-05 4.7E-04 1.7E-06 2.1E-05 4.0E-05 4.9E-05 5.2E-04 2.9E-04 3.7E-05 2.4E-05

1980 2.3E-05 2.6E-04 1.1E-06 2.9E-05 4.8E-05 4.5E-05 4.5E-04 4.4E-05 1.9E-05

1981 2.6E-05 6.4E-05 1.4E-04 2.5E-05 2.9E-05 3.3E-05 3.2E-04 4.6E-05 2.0E-05

1982 6.1E-05 4.0E-05 6.2E-05 2.0E-05 3.5E-05 7.4E-05 2.2E-04 5.3E-05 2.7E-05

1983 3.1E-05 6.5E-05 9.2E-05 2.3E-05 3.2E-05 3.3E-05 2.1E-04 1.8E-04 2.3E-05

1984 1.9E-05 1.0E-04 2.3E-04 3.8E-05 2.0E-05 3.8E-05 1.0E-03 5.6E-05 1.7E-05

1985 1.5E-05 5.8E-05 2.1E-04 3.1E-05 2.7E-05 2.7E-05 1.3E-03 5.2E-05 2.3E-05

1986 1.5E-04 6.2E-05 5.5E-04 2.8E-05 1.9E-04 4.8E-05 2.8E-04 2.5E-05 3.2E-05

1987 5.5E-05 2.3E-05 2.8E-03 1.7E-05 2.5E-05 1.6E-05 1.1E-04 3.2E-05 1.6E-05

1988 9.2E-05 2.8E-05 2.1E-04 2.8E-05 2.2E-05 1.5E-05 5.1E-05 1.2E-05 1.6E-05

1989 8.6E-04 5.8E-05 5.4E-04 5.0E-05 8.0E-05 2.7E-04 3.5E-04 5.4E-04 6.8E-04

1990 9.9E-05 3.2E-05 3.5E-04 3.1E-05 3.7E-05 4.1E-05 5.0E-04 4.3E-05 1.1E-04 1.0E-05

1991 3.1E-03 2.1E-05 3.8E-04 3.5E-05 7.2E-05 3.2E-05 3.5E-04 5.4E-07 2.2E-04 1.4E-05

1992 1.5E-04 1.1E-04 2.3E-03 8.0E-05 7.3E-05 3.9E-04 8.8E-04 2.4E-04 8.8E-05 3.8E-05

1993 2.0E-04 4.9E-05 4.0E-04 8.0E-05 9.0E-05 7.8E-05 2.7E-03 1.7E-04 4.2E-04 8.4E-04

1994

1995 4.1E-05 1.1E-05 2.4E-04 1.5E-05 5.0E-05 1.2E-05 3.9E-04 3.8E-05 4.0E-05 1.1E-05

1996 7.5E-04 3.0E-04 1.6E-05 4.6E-05 4.5E-04 6.1E-04 5.8E-05 1.9E-05 3.7E-06

1997 1.1E-05 2.6E-05 1.1E-05 1.5E-06

1998 4.7E-04 6.7E-05 9.5E-05 1.1E-05 2.7E-05 7.4E-04 9.5E-06

1999 5.9E-05

2000 8.5E-05 2.7E-04 5.9E-05 2.1E-05 4.1E-04 3.4E-05 2.8E-03

2001 4.5E-04 1.2E-04 1.2E-04 1.9E-04 6.8E-05 9.6E-05 1.7E-05 5.0E-04 8.5E-06

Year

Area Site 

averageb

Site 

maximumc 15 16 18 19 20 23 25 27 28

1971 1.9E-04 8.0E-05 2.3E-04 9.4E-05 8.4E-05 3.9E-05 2.26E-04 7.21E-04

1972 1.7E-04 1.5E-04 2.7E-03 3.5E-04 1.7E-04 5.1E-05 7.63E-04 4.29E-03

1973 4.4E-05 3.8E-05 7.2E-05 5.6E-05 3.7E-05 3.3E-05 1.46E-04 8.59E-04

1974 6.0E-05 6.6E-05 8.5E-05 8.6E-05 7.2E-05 1.2E-04 8.64E-05 2.11E-04

1975 4.3E-05 3.4E-05 4.4E-05 5.7E-05 6.8E-05 7.0E-05 6.27E-05 1.68E-04

1976 1.6E-04 9.9E-05 5.7E-04 2.0E-04 1.4E-04 6.3E-05 4.82E-04 3.18E-03

1977 2.6E-05 2.3E-05 2.7E-05 3.3E-05 3.1E-05 2.3E-05 6.93E-05 2.55E-04

1978 5.1E-05 6.2E-05 5.8E-05 7.0E-05 9.4E-05 3.9E-05 1.33E-04 5.42E-04

1979 5.2E-05 2.1E-05 2.0E-05 2.2E-05 1.8E-05 1.6E-05 2.3E-05 9.98E-05 5.24E-04

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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Year

Area

1 2 3 4 5 6 7 9 10 11 12

1980 6.4E-05 9.5E-05 3.5E-05 3.1E-05 2.1E-05 1.6E-05 7.88E-05 4.52E-04

1981 9.3E-05 2.6E-05 2.3E-05 2.9E-05 2.4E-05 2.5E-05 1.6E-05 5.88E-05 3.22E-04

1982 4.2E-05 1.7E-05 2.2E-05 2.1E-05 1.7E-05 5.04E-05 2.15E-04

1983 2.2E-05 1.5E-05 3.0E-05 2.8E-05 3.0E-05 5.85E-05 2.14E-04

1984 3.2E-05 2.8E-05 2.5E-05 2.9E-05 3.2E-05 4.8E-05 1.15E-04 1.02E-03

1985 4.0E-05 2.1E-05 3.5E-05 2.2E-05 2.6E-05 3.1E-05 1.30E-04 1.33E-03

1986 3.9E-05 2.8E-05 2.2E-05 6.6E-05 2.1E-05 2.3E-05 1.03E-04 5.45E-04

1987 1.8E-05 1.4E-05 1.7E-05 1.6E-05 2.3E-05 2.13E-04 2.81E-03

1988 2.4E-05 1.4E-05 3.5E-05 1.5E-05 1.5E-05 4.15E-05 2.14E-04

1989 2.3E-05 1.1E-05 9.0E-06 1.2E-04 9.5E-06 4.8E-06 2.40E-04 8.55E-04

1990 8.2E-05 5.8E-06 3.9E-06 8.5E-05 1.1E-05 3.5E-05 9.17E-05 4.98E-04

1991 1.2E-04 2.9E-05 8.8E-06 1.3E-05 2.6E-05 7.7E-06 2.77E-04 3.10E-03

1992 1.1E-03 1.9E-05 8.9E-06 1.3E-05 1.8E-05 1.6E-04 3.52E-04 2.26E-03

1993 3.3E-04 1.3E-05 7.7E-06 6.3E-06 3.4E-06 1.1E-05 3.37E-04 2.70E-03

1994

1995 1.6E-04 1.4E-05 1.0E-06 7.4E-06 4.5E-06 6.92E-05 3.90E-04

1996 2.4E-04 1.8E-06 3.5E-06 4.5E-06 6.3E-06 4.2E-06 2.6E-06 1.57E-04 7.50E-04

Year

Area Site 

averageb

Site 

maximumc 15 16 18 19 20 23 25 27 28

1997 2.9E-05 1.9E-07 2.0E-05 2.0E-05 6.0E-07 1.0E-06 1.6E-06 1.11E-05 2.90E-05

1998 6.3E-05 3.1E-05 3.3E-05 1.2E-05 6.3E-06 1.30E-04 7.35E-04

1999 1.1E-05 1.1E-05 2.0E-05 6.2E-06 2.14E-05 5.90E-05

2000 2.6E-04 1.0E-05 6.6E-06 1.1E-05 3.63E-04 2.83E-03

2001 1.4E-05 9.6E-06 1.7E-05 7.6E-06 2.0E-06 4.4E-06 1.02E-04 5.04E-04

a. Source: NTS environmental reports in References section.

b. Values represent the arithmetic average of the area average concentrations for 1971 through 1988 and the arithmetic average of the 

area maximum concentrations for 1989 through 2001.

c. Values represent the maximum of the average area concentrations for 1971 through 1988 and the maximum of the maximum area 

concentrations for 1989 through 2001.

Table A-2. Annual inhalation intakes from 239Pu for sampled areas (Bq). 

Year

Area

1 2 3 4 5 6 7 9 10 11 12

1971 0.0329 0.0587 0.0147 0.0105 0.0153 0.0640 0.0185 0.0081 0.0145

1972 0.0131 0.0199 0.0327 0.0124 0.1190 0.3810 0.0258 0.0216 0.0698

1973 0.0077 0.0198 0.0183 0.0078 0.0107 0.0763 0.0039 0.0216 0.0038

1974 0.0071 0.0060 0.0112 0.0044 0.0070 0.0187 0.0051 0.0071 0.0052

1975 0.0026 0.0034 0.0130 0.0038 0.0056 0.0149 0.0051 0.0044 0.0029

1976 0.0255 0.0153 0.0924 0.0133 0.0392 0.2824 0.0189 0.0210 0.0248

1977 0.0024 0.0058 0.0226 0.0030 0.0049 0.0225 0.0041 0.0107 0.0019

1978 0.0042 0.0118 0.0250 0.0054 0.0088 0.0481 0.0258 0.0095 0.0055

1979 0.0066 0.0413 0.0001 0.0018 0.0035 0.0044 0.0465 0.0258 0.0033 0.0021

1980 0.0020 0.0230 0.0001 0.0026 0.0043 0.0040 0.0401 0.0039 0.0017

1981 0.0023 0.0056 0.0124 0.0022 0.0026 0.0029 0.0286 0.0041 0.0018

1982 0.0054 0.0036 0.0055 0.0018 0.0031 0.0066 0.0191 0.0047 0.0024

1983 0.0028 0.0057 0.0082 0.0020 0.0029 0.0029 0.0190 0.0162 0.0020

1984 0.0017 0.0088 0.0202 0.0034 0.0018 0.0034 0.0906 0.0050 0.0015

1985 0.0013 0.0051 0.0189 0.0027 0.0024 0.0024 0.1181 0.0046 0.0020

1986 0.0131 0.0055 0.0484 0.0025 0.0165 0.0043 0.0249 0.0022 0.0028

1987 0.0048 0.0020 0.2495 0.0015 0.0022 0.0014 0.0098 0.0028 0.0014

1988 0.0082 0.0025 0.0190 0.0025 0.0020 0.0013 0.0045 0.0011 0.0014

1989 0.0759 0.0052 0.0478 0.0045 0.0071 0.0240 0.0311 0.0480 0.0604

1990 0.0088 0.0029 0.0308 0.0027 0.0033 0.0036 0.0442 0.0038 0.0095 0.0009

1991 0.2753 0.0018 0.0340 0.0031 0.0064 0.0029 0.0311 0.0000 0.0199 0.0013

1992 0.0133 0.0098 0.2007 0.0071 0.0065 0.0346 0.0781 0.0213 0.0078 0.0034

1993 0.0180 0.0044 0.0352 0.0071 0.0079 0.0069 0.2398 0.0151 0.0373 0.0746

1994

1995 0.0036 0.0010 0.0216 0.0014 0.0045 0.0011 0.0346 0.0033 0.0036 0.0010

1996 0.0666 0.0000 0.0268 0.0015 0.0041 0.0400 0.0542 0.0051 0.0017 0.0003

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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1997 0.0010 0.0023 0.0010 0.0001

1998 0.0416 0.0060 0.0084 0.0010 0.0024 0.0653 0.0008

1999

2000 0.0075 0.0239 0.0052 0.0018 0.0366 0.0030 0.2509

2001 0.0404 0.0011 0.0103 0.0166 0.0061 0.0086 0.0015 0.0447 0.0008

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1971 0.0171 0.0071 0.0201 0.0083 0.0075 0.0035 0.0200 0.0640

1972 0.0150 0.0129 0.2424 0.0309 0.0147 0.0045 0.0677 0.3810

1973 0.0039 0.0034 0.0063 0.0050 0.0033 0.0029 0.0130 0.0763

1974 0.0053 0.0059 0.0075 0.0076 0.0064 0.0104 0.0077 0.0187

1975 0.0038 0.0030 0.0039 0.0051 0.0060 0.0062 0.0056 0.0149

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1976 0.0139 0.0088 0.0504 0.0178 0.0122 0.0056 0.0428 0.2824

1977 0.0023 0.0020 0.0024 0.0029 0.0028 0.0020 0.0062 0.0226

1978 0.0045 0.0055 0.0052 0.0062 0.0083 0.0035 0.0118 0.0481

1979 0.0046 0.0019 0.0017 0.0020 0.0016 0.0014 0.0020 0.0089 0.0465

1980 0.0057 0.0085 0.0031 0.0027 0.0019 0.0014 0.0070 0.0401

1981 0.0083 0.0023 0.0020 0.0025 0.0022 0.0022 0.0014 0.0052 0.0286

1982 0.0037 0.0015 0.0019 0.0019 0.0015 0.0045 0.0191

1983 0.0020 0.0013 0.0026 0.0025 0.0027 0.0052 0.0190

1984 0.0028 0.0025 0.0022 0.0025 0.0028 0.0043 0.0102 0.0906

1985 0.0036 0.0019 0.0031 0.0020 0.0023 0.0028 0.0115 0.1181

1986 0.0035 0.0025 0.0020 0.0058 0.0018 0.0020 0.0092 0.0484

1987 0.0016 0.0012 0.0015 0.0014 0.0020 0.0000 0.0189 0.2495

1988 0.0022 0.0012 0.0031 0.0013 0.0013 0.0037 0.0190

1989 0.0021 0.0010 0.0008 0.0104 0.0008 0.0004 0.0213 0.0759

1990 0.0073 0.0005 0.0003 0.0075 0.0010 0.0031 0.0081 0.0442

1991 0.0109 0.0026 0.0008 0.0012 0.0023 0.0007 0.0246 0.2753

1992 0.0977 0.0017 0.0008 0.0011 0.0016 0.0142 0.0312 0.2007

1993 0.0293 0.0012 0.0007 0.0006 0.0003 0.0010 0.0300 0.2398

1994 0.0000 0.0000

1995 0.0142 0.0012 0.0001 0.0007 0.0004 0.0061 0.0346

1996 0.0213 0.0002 0.0003 0.0000 0.0004 0.0006 0.0004 0.0002 0.0124 0.0666

1997 0.0026 0.0000 0.0018 0.0018 0.0001 0.0001 0.0001 0.0010 0.0026

1998 0.0056 0.0027 0.0029 0.0010 0.0006 0.0115 0.0653

1999 0.0010 0.0010 0.0018 0.0006 0.0011 0.0018

2000 0.0229 0.0009 0.0006 0.0010 0.0322 0.2509

2001 0.0012 0.0008 0.0015 0.0007 0.0002 0.0004 0.0090 0.0447

Some covered employees remained on the site continuously for weeks at a time. However, because 

most of the nonworking hours were spent indoors where ambient air particulate loadings would be 

much less than the outdoor loadings and because of the conservative assumptions that were used to 

estimate the values in Table A-2, adjustment of the tabular data is not required to ensure intakes are 

not underestimated for these individuals. In addition, employees who lived on the site during their 

work week would have been housed in Area 12 or Area 23 (Mercury). For most years the values in 

Table A-1 for these locations are less than the site average values.

It is assumed that plutonium could be any of absorption type S, Super S, or M, depending on which 

type delivers the maximum organ dose. 

A.3 Soil Contamination and Decay Corrections

Extensive studies were performed in the 1980s to quantify residual contamination at NTS (McArthur 

and Kordas 1983, 1985; McArthur and Mead 1987, 1988, 1989; McArthur 1991). Table A-3 lists the 

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results of these studies (McArthur 1991). Table A-4 lists the total areal depositions based on the 

inventory values in Table A-3 divided by the respective areal size. The results in Table A-4 are 

representative of areas of NTS that have been shown to contain measurable levels of contamination. 

These areas actually represent only about one-third of the total area within the boundaries of NTS.

Because the data in Table A-3 are representative of soil contamination in 1991, these values were 

decay-corrected to the beginning of 1963 – the first year after the cessation of atmospheric testing. 

Table A-5 lists the decay-corrected areal soil deposition.

Table A-3. Inventory of contaminated soil (Ci). 

Area Area (mi2) Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 26.5 4.2 6.5 24 1.1 8.8 15 15 0.1 0.5

2 19.7 2.9 8.6 22 1.2 24 46 14 0.4

3 32.3 4.6 3.1 37 1 12 33 18 0.1 0.5

4 16 6.6 13 40 1.6 12 13 9.1 0.2

5 2.9 0.6 0.1 4.8 0.6 0.4 0.9 10 0.2 0

6 32.3 1.7 3.3 8.4 0.2 2.8 3.5 0

7 19.3 2.2 0.6 16 1 5.2 9.2 22 0.2 0.3

8 13.9 17 8 110 5.7 42 25 4.4 0.6

9 20 4.2 2.2 89 0.7 8.7 13 23 0.2 0.3

10 20 19 19 110 9.7 84 55 2.2 0.3 5

11 4 3.3 0.5 29 0 0.5 0.3

12 39.6 5.7 8.5 39 1.2 20 17

15 35.3 8 7.8 63 0.3 19 22

16 14.3 0.7 1.5 3.7 0.1 2.9 3.7

17 31.4 2.8 4.5 18 1.0 15 19

18 27.3 19 5.6 100 0.7 10 17 1.1 0.1 0.8

19 148.3 21 32 140 1.1 36 31

20 6.2 23 30 41 7.9 5.5 4.3 13 1.6 4.8

25 0.9 0.2 0.1 0.4

26 0.2

30 0.03 3.2 4.5 14 0.8 1.5 1.3 0.7 0.1 0.2

Source: McArthur (1991).

Table A-4. Radionuclide areal soil deposition (Bq/m2).

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 2.26E+03 3.50E+03 1.29E+04 5.93E+02 4.74E+03 8.09E+03 8.09E+03 5.39E+01 2.70E+02

2 2.10E+03 6.24E+03 1.60E+04 8.70E+02 1.74E+04 3.34E+04 1.02E+04 0.00E+00 2.90E+02

3 2.03E+03 1.37E+03 1.64E+04 4.42E+02 5.31E+03 1.46E+04 7.96E+03 4.42E+01 2.21E+02

4 5.89E+03 1.16E+04 3.57E+04 1.43E+03 1.07E+04 1.16E+04 8.13E+03 0.00E+00 1.79E+02

5 2.96E+03 4.93E+02 2.36E+04 2.96E+03 1.97E+03 4.43E+03 4.93E+04 9.85E+02 0.00E+00

6 7.52E+02 1.46E+03 3.72E+03 8.85E+01 1.24E+03 1.55E+03 0.00E+00 0.00E+00 0.00E+00

7 1.63E+03 4.44E+02 1.18E+04 7.40E+02 3.85E+03 6.81E+03 1.63E+04 1.48E+02 2.22E+02

8 1.75E+04 8.22E+03 1.13E+05 5.86E+03 4.32E+04 2.57E+04 4.52E+03 0.00E+00 6.17E+02

9 3.00E+03 1.57E+03 6.36E+04 5.00E+02 6.21E+03 9.29E+03 1.64E+04 1.43E+02 2.14E+02

10 1.36E+04 1.36E+04 7.86E+04 6.93E+03 6.00E+04 3.93E+04 1.57E+03 2.14E+02 3.57E+03

11 1.18E+04 1.79E+03 1.04E+05 0.00E+00 1.79E+03 1.07E+03 0.00E+00 0.00E+00 0.00E+00

12 2.06E+03 3.07E+03 1.41E+04 4.33E+02 7.22E+03 6.13E+03 0.00E+00 0.00E+00 0.00E+00

15 3.24E+03 3.16E+03 2.55E+04 1.21E+02 7.69E+03 8.90E+03 0.00E+00 0.00E+00 0.00E+00

16 6.99E+02 1.50E+03 3.70E+03 9.99E+01 2.90E+03 3.70E+03 0.00E+00 0.00E+00 0.00E+00

17 1.27E+03 2.05E+03 8.19E+03 4.55E+02 6.82E+03 8.64E+03 0.00E+00 0.00E+00 0.00E+00

18 9.94E+03 2.93E+03 5.23E+04 3.66E+02 5.23E+03 8.90E+03 5.76E+02 5.23E+01 4.19E+02

19 2.02E+03 3.08E+03 1.35E+04 1.06E+02 3.47E+03 2.99E+03 0.00E+00 0.00E+00 0.00E+00

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20 5.30E+04 6.91E+04 9.45E+04 1.82E+04 1.27E+04 9.91E+03 3.00E+04 3.69E+03 1.11E+04

25 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.17E+03 1.59E+03 6.35E+03 0.00E+00 0.00E+00

26 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

30 1.52E+05 2.14E+05 6.67E+05 3.81E+04 7.14E+04 6.19E+04 3.33E+04 4.76E+03 9.52E+03

Table A-5. Radionuclide areal soil deposition (Bq/m2) decay, corrected to 1963. Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 2.37E+03 4.37E+03 1.29E+04 2.35E+04 9.06E+03 1.57E+04 3.47E+04 4.89E+02 1.35E+04

2 2.20E+03 7.78E+03 1.60E+04 3.45E+04 3.32E+04 6.49E+04 4.35E+04 0.00E+00 1.45E+04

3 2.13E+03 1.71E+03 1.76E+04 1.01E+04 2.84E+04 3.41E+04 4.01E+02 1.11E+04

4 6.16E+03 1.45E+04 3.57E+04 5.67E+04 2.05E+04 2.26E+04 3.48E+04 8.93E+03

5 3.09E+03 6.15E+02 2.37E+04 1.17E+05 3.76E+03 8.63E+03 2.11E+05 8.94E+03 0.00E+00

6 7.86E+02 1.82E+03 3.72E+03 3.51E+03 2.36E+03 3.01E+03

7 1.70E+03 5.54E+02 1.19E+04 2.94E+04 7.35E+03 1.33E+04 6.98E+04 1.34E+03 1.11E+04

8 1.83E+04 1.03E+04 1.13E+05 2.33E+05 8.24E+04 5.00E+04 1.94E+04 3.08E+04

9 3.14E+03 1.96E+03 6.36E+04 1.98E+04 1.19E+04 1.81E+04 7.04E+04 1.30E+03 1.07E+04

10 1.42E+04 1.69E+04 7.86E+04 2.75E+05 1.15E+05 7.65E+04 6.74E+03 1.94E+03 1.79E+05

11 1.23E+04 2.23E+03 1.04E+05 0.00E+00 3.41E+03 2.09E+03

12 2.15E+03 3.83E+03 1.41E+04 1.72E+04 1.38E+04 1.19E+04

15 3.39E+03 3.94E+03 2.55E+04 4.82E+03 1.47E+04 1.73E+04

16 7.31E+02 1.87E+03 3.70E+03 3.97E+03 5.53E+03 7.20E+03

17 1.33E+03 2.55E+03 8.20E+03 1.81E+04 1.30E+04 1.68E+04

18 1.04E+04 3.66E+03 5.24E+04 1.45E+04 9.99E+03 1.73E+04 2.47E+03 4.75E+02 2.09E+04

19 2.12E+03 3.85E+03 1.35E+04 4.21E+03 6.62E+03 5.81E+03

20 5.54E+04 8.62E+04 9.45E+04 7.23E+05 2.42E+04 1.93E+04 1.28E+05 3.34E+04 5.53E+05

25 6.06E+03 3.09E+03 2.72E+04

26

30 1.59E+05 2.67E+05 6.67E+05 1.51E+06 1.36E+05 1.21E+05 1.43E+05 4.32E+04 4.76E+05

A.4 Scaling Factors for Inhalation Intake Estimates

Because the air sampling program did not provide isotopic analyses for all identified radionuclides in 

NTS soils, scaling factors were developed to estimate potential intakes for these radionuclides based 

on their relative abundances in comparison with 239Pu with soil contamination data (McArthur 1991) 

decay-corrected to 1963. These area-specific ratios are listed in Table A-6.

The scaling factors in Table A-6 were used in conjunction with the 239Pu intakes in Table A-2 to 

determine potential environmental intakes of all radionuclides important to dose (Table A-6) that have 

been identified as persistent in NTS soils. For dose reconstruction, maximum intakes of these 

radionuclides were calculated by selecting the maximum annual intake of plutonium (i.e., 0.4131 Bq/yr 

derived for Area 9 in 1972) and multiplying this value by the maximum scaling factor for each of the 

radionuclides in Table A-6. 

A.5 Correction for Resuspension for Early Times After Atmospheric Tests

Anspaugh et al. (2002) stated that, based on empirical observations, concentration of resuspended 

radionuclides in air have been noted to display a strong time dependence early after deposition and 

that this pathway might be important for reoccupation of contaminated property. Anspaugh et al. also 

stated that there has not been universal agreement that resuspension is an important pathway but 

that it is now generally accepted that there are a few instances in which the pathway could be the 

dominant one. Many observations have shown that the rate of resuspension decreases rapidly with 

time, and that for accident situations resuspension is only of importance (in comparison with the 

inhalation exposure from the initial cloud passage) over short periods. For this reason, Anspaugh et 

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al. (2002) stated that the resuspension factor model has been widely used to predict the concentration 

of resuspended radionuclides early after initial deposition while the mass loading model (which uses 

measurements of dust loading in air and soil contamination data to predict air concentration of 

Table A-6. Abundance of radionuclides in NTS soils relative to 239Pu decay, corrected to 1963. 

Area Am-241 Pu-238 Pu-239, 240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 0.183 0.338 1.000 1.818 0.700 1.216 2.677 0.038 1.041

2 0.138 0.487 1.000 2.164 2.081 4.067 2.726 0.908

3 0.130 0.104 1.000 1.072 0.619 1.735 2.084 0.024 0.675

4 0.172 0.405 1.000 1.587 0.572 0.632 0.975 0.250

5 0.131 0.026 1.000 4.958 0.159 0.365 8.925 0.378

6 0.212 0.490 1.000 0.944 0.636 0.810

7 0.144 0.047 1.000 2.479 0.620 1.118 5.890 0.113 0.937

8 0.162 0.091 1.000 2.055 0.728 0.442 0.171 0.273

9 0.049 0.031 1.000 0.312 0.187 0.284 1.107 0.020 0.168

10 0.181 0.215 1.000 3.498 1.457 0.973 0.086 0.025 2.271

11 0.119 0.021 1.000 0.033 0.020

12 0.153 0.272 1.000 1.220 0.978 0.848

15 0.133 0.154 1.000 0.189 0.575 0.679

16 0.198 0.505 1.000 1.072 1.495 1.945

17 0.163 0.312 1.000 2.204 1.590 2.053

18 0.199 0.070 1.000 0.278 0.191 0.331 0.047 0.009 0.400

19 0.157 0.285 1.000 0.312 0.491 0.431

20 0.586 0.912 1.000 7.643 0.256 0.204 1.358 0.354 5.849

25

26

30 0.229 0.321 1.000 0.057 0.107 0.093 0.050 0.007 0.014

Maximum scaling factor 0.586 0.912 1.000 7.64 2.08 4.07 8.93 0.378 5.85

Scaled maximum intake, 

Bq/yr

0.223 0.347 0.381 2.91 0.792 1.55 3.40 0.144 2.23

radionuclides) has generally been preferred for times long after deposition. However, Anspaugh et al.

also stated that it is always preferable to rely on actual measurements that are performed over long 

periods (such as those in this attachment).

Anspaugh et al. (2002) presented several resuspension models that have been proposed but 

concluded that they can be over- or under-predictive at various times after deposition in comparison 

with empirical observations. However, with expanded datasets from Hicks (1981e) and others in the 

1980s, Anspaugh et al. proposed a resuspension model that more accurately describes the observed 

results over the entire timespan of the expanded dataset for NTS:

5 0.07 9 0.003 9 1 1 [10 6 10 10 ] 10 m t t Se e f

− − −− − ± − = +× + × (A-1)

A depiction of Equation A-1 is provided in Figure A-1 from time t equal zero to 1,000 days after 

detonation. As Figure A-1 shows, the resuspension factor Sf ranges from about 10-5 early after 

deposition, falls rapidly during the first 100 or so days to a value of about 10-8, and then approaches a 

value of 10-9 after a few years. The factor of 10 at the end of the equation is a statement of 

uncertainty in the model. If the mass loading approach is more predictive at times long after initial 

deposition and the resuspension proposed by Anspaugh et al. (2002) is predictive of the observed 

results over the expanded dataset (including those developed in the 1980s), the factor of 10-9 could be 

taken to be the resuspension factor that would be predictive of the mass loading process that is 

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thought to be more important during the times when air monitoring data are available (i.e., 1971 

through 2001; see Section A.2). When the Anspaugh et al. proposed resuspension model (i.e., 

Equation A-1) is integrated from 180 to 545 days and compared to the result of the integral of the 

constant 10-9 over the same period, a factor is developed that can be used to correct the intakes 

derived from air sampling data (i.e., times long after initial deposition) for the early resuspension 

phenomenon that has been observed at NTS. The period of integration was selected to begin at day 

180 because the last atmospheric tests at NTS were in July 1962 and, therefore, about 180 days had 

passed before the beginning of 1963 (the starting date for reconstruction of internal dose at NTS in 

the absence of bioassay results for the employee). For the Anspaugh et al. model, the early 

resuspension correction factor has been determined to be 3.12. In a similar manner, correction 

factors for 1964 and 1965 were determined to be 1.72 and 1.24, respectively. Therefore, for dose 

reconstruction, the scaled maximum inhalation intakes in Table A-6 should be multiplied by these 

factors to account for early resuspension from 1963 through 1965. These resultant intakes are listed 

in Table A-7. If necessary, these intakes can be prorated for time less than a year for a best estimate 

if the worker was on the site for only a fraction of the year.

Figure A-1. Resuspension factor as a function of time after initial deposition. 

Table A-7. Scaled inhalation intakes corrected for early resuspension in 1963 (Bq/yr). 

Year of intake Am-241 Pu-238 Pu-239, 240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1963 0.70 1.08 1.19 9.08 2.47 4.84 10.61 0.45 6.96

1964 0.38 0.59 0.65 4.98 1.35 2.65 5.81 0.25 3.81

1965 0.28 0.43 0.47 3.61 0.98 1.92 4.22 0.18 2.77

All subsequent years 0.223 0.347 0.381 2.91 0.792 1.55 3.40 0.144 2.23

Resuspension Factor

1.3 E-09 m-1

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

0 100 200 300 400 500 600 700 800 900 1000

Days

meter-1

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A.6 Corrections for Inhalation Dose from Short-lived Fission and Activation Products

During the early and mid-1960s, workers could have been exposed to fallout from atmospheric testing 

that contained short-lived fission and activation products (144Ce, 106Ru, etc.) that have not persisted in 

NTS soils in measurable quantities. For dose reconstruction, two methods are described to account 

for possible exposure to these radionuclides. 

Potential inhalation dose from short-lived fission and activation products can be estimated using data 

developed by Hicks (1981e). Hicks provided estimates of ground deposition of 152 fission products 

and 25 neutron-induced nuclides from nuclear weapons tests. The data for these calculations were 

for the last atmospheric tests NTS conducted in 1962 as a function of time after detonation. 

Spreadsheets were developed to multiply the relative abundance of each of these radionuclides by its 

associated International Commission on Radiological Protection (ICRP) Publication 68 (ICRP 1995) 

inhalation organ dose conversion factor (Bunker 1999) to determine the relative importance of each of 

the 177 radionuclides to total organ dose. With the exception of 90Sr, the inhalation dose conversion 

factors were for the absorption type (i.e., S, M, or F) that produced the largest dose to the organ of 

interest. For 90Sr, only absorption type F was used. In addition, because ICRP Publication 68 organ 

dose factors represent the 50-year committed dose, the IMBA computer program was used to develop 

the annual organ-specific dose conversion factor for 241Am, which was used in the calculations to 

better represent the relative importance of 241Am, which delivers dose to the affected organ over long 

periods. Similarly, annual organ-specific dose conversion factors were developed for 90Sr to provide a 

more accurate time-dependant correction factor.

The development of the short-lived fission and activation product correction factors based on the 

Hicks data must be adjusted for fractionation. Fractionation is a phenomenon due to chemical and 

physical separation of the radionuclides in the fireball in first few minutes after detonation. Within the 

first minute after detonation, the vaporized soil components condense with other refractory elements 

and begin to fall to the surface. The volatile elements (except krypton and xenon) and their progeny 

condense in 6 to 8 minutes and begin falling to the surface (Hicks 1981e). Because the Hicks data 

were developed to estimate offsite levels of fallout and resultant dose, fractionation effects were 

simulated in these data by the removal of a fraction of the refractory nuclides from the calculated 

abundances. In general, air drops were assumed to be unfractionated and offsite fallout from surface 

and cratering tests was assumed to have 0.4 of the refractory elements. For all other types of tests, 

offsite fallout was assumed to have 0.5 of the refractory elements present. Therefore, the refractory 

elements (beryllium, sodium, manganese, iron, cobalt, copper, yttrium, zirconium, niobium, barium, 

rare earths, thorium, uranium, neptunium, plutonium, americium, and curium) in the Hicks data must 

be adjusted to produce the best estimate of their enriched abundances in the onsite environment to 

which workers could have been exposed. Adjustment factors for each radionuclide were determined 

from data in Hicks (1984); this report provided relative abundances of radionuclides assuming no 

fraction, 50% fraction, and 90% fraction of refractory elements. From these data, ratios were 

developed for the 50% fractionation case (Table A-8). These ratios were used to deplete the 

refractory elements in the far-field (i.e., offsite) environment to estimate doses to offsite individuals. 

Therefore, to enrich the near-field (i.e., onsite) environment, the inverse of these ratios was applied to 

the Hicks SMALL BOY data (see below). These inverse ratios were applied twice because the Hicks 

SMALL BOY data were provided to estimate fallout in the offsite environment. The first application 

results in the data that represent no fractionation while the second application results in data that are 

enriched with refractory elements.

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In Table A-8, the radionuclides with ratios less than 1.0 represent the radionuclides that condense as 

refractory elements while the radionuclides with ratios greater than 1.0 are assumed to behave as 

volatile elements. Although strontium is a refractory element, for these calculation, the assumption is 

that it behaves like a volatile element because its fission precursors are noble gases (i.e., volatile 

elements). 

Table A-8. Far-field refractory enrichment ratios. Radionuclide Ratio

Radionuclide RatioRadionuclide RatioRadionuclide RatioRadionuclide RatioRadionuclide Ratio

Be-7 0.68 Br-82 1.36 Tc-101 1.36 Cd-118 1.36 I-132 1.36 Pr-145 0.68

Na-24 0.68 Se-83 1.36 Mo-102 1.36 In-118 1.36 Te-133m 1.36 Ce-146 0.68

Mn-54 0.68 Br-83 1.36 Tc-102m 1.36 Cd-119 1.36 Te-133 1.36 Pr-146 0.68

Fe-55 0.68 Kr-83m 1.37 Tc-102 1.36 In-119m 1.36 I-133 1.36 Pr-147 0.68

Fe-59 0.68 Br-84 1.36 Ru-103 1.36 In-119 1.36 Xe-133m 1.36 Nd-147 0.68

Cu-64 0.68 Kr85m 1.36 Rh-103m 1.36 Sn-121 1.36 Xe-133 1.36 Nd-149 0.68

Cu-67 0.68 Kr-87 1.36 Tc-104 1.36 Sn-123m 1.37 Te-134 1.36 Pm-149 0.68

W-181 1.36 Kr-88 1.36 Ru-105 1.36 Sn-123 1.36 I-134 1.36 Pm-150 0.68

W-185 1.35 Rb-88 1.36 Rh-105m 1.36 Sn-125 1.36 I-135 1.36 Nd-151 0.68

W-187 1.36 Rb-89 1.36 Rh-105 1.36 Sb-125 1.36 Xe-135m 1.36 Pm-151 0.68

W-188 1.36 Sr-89 1.36 Ru-106 1.36 Sb-126 1.36 Xe-135 1.36 Pm-152 0.68

U-237 0.68 Sr-90 1.36 Rh-106 1.36 Sn-127 1.36 Cs-136 1.36 Sm-153 0.68

U-239 0.68 Sr-91 1.19 Rh-107 1.36 Sb-127 1.36 Cs-137 1.36 Sm-155 0.68

U-240 0.68 Y-91m 1.19 Pd-107m 1.36 Te-127 1.36 Ba-137m 1.36 Eu-155 0.68

Np-239 0.68 Y-91 1.19 Pd-109 1.36 Sn-128 1.36 Xe-138 1.36 Sm-156 0.68

Np240m 0.69 Sr-92 0.68 Ag-109m 1.36 Sb-128m 1.37 Cs-138 1.36 Eu-156 0.68

Np-240 0.68 Y-92 0.68 Pd-111m 1.37 Sb-128 1.37 Cs-139 1.36 Eu-157 0.68

Am-241 0.68 Sr-93 0.68 Pd-111 1.36 Sn-129m 1.36 Ba-139 1.36 Eu-158 0.68

Cm-242 0.68 Y-93 0.68 Ag-111m 1.36 Sn-129 1.36 Ba-140 1.16 Eu-159 0.68

Ge-75 1.36 Y-94 0.68 Ag-111 1.36 Sb-129 1.36 La-140 1.16 Gd-159 0.68

Ge-77 1.36 Y-95 0.68 Pd-112 1.36 Te-129m 1.36 Ba-141 0.90 Tb-161 0.68

As-77 1.36 Zr-95 0.68 Ag-112 1.36 Te-129 1.36 La-141 0.90

Se-77m 1.36 Nb-95 0.68 Ag-113 1.36 Sb-130m 1.36 Ce-141 0.90

Ge-78 1.36 Zr-97 0.68 Ag-115 1.36 Sb-130 1.36 Ba-142 0.68

As-78 1.36 Nb-97m 0.68 Cd-115m 1.36 I-130 1.36 La-142 0.68

As-79 1.37 Nb-97 0.68 Cd-115 1.36 Sb-131 1.36 La-143 0.68

Se-79m 1.36 Nb-98 0.68 In-115m 1.37 Te-131m 1.36 Ce-143 0.68

Br-80 1.36 Mo-99 0.68 Cd-117 1.36 Te-131 1.36 Pr-143 0.68

Se-81m 1.36 Tc-99m 0.68 In-117m 1.36 I-131 1.36 Ce-144 0.68

Se-81 1.36 Mo-101 1.36 In-117 1.36 Te-132 1.36 Pr-144 0.68

The fission and activation product correction factor was developed based on the relative contribution 

of 90Sr to total organ dose. Strontium-90 was chosen because Hicks (1984) provided time-dependant 

abundances for this radionuclide and 90Sr continues to persist in the NTS environment. Therefore, by 

determining the relative importance of organ dose from 90Sr to the total dose from all 177 

radionuclides, a multiplication factor was developed, by which the various organ doses from the 90Sr 

intakes (Table A-7) can be multiplied, to account for short-lived fission and activation products. For 

example, using the Hicks data for STORAX SMALL BOY, which was the next-to-last atmospheric test 

at NTS on July 14, 1962 (STORAX LITTLE FELLER I was the last atmospheric test on July 17, 1962), 

it was determined that the relative importance of 90Sr dose to the lung to the total dose from all 177 

radionuclides varied, in a mostly linear fashion, from 0.00284 to 0.0589 from 1 to 365 days after 

detonation, respectively (see Figure A-2). Using the trend line function, an expression for the relative 

importance of 90Sr dose y to total dose as a function of the number of days after detonation x was 

developed: 

y x = + 0.0001 0.0074 (A-2)

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 70 of 88

To ensure favorability to claimants, the relative importance of 90Sr lung dose to the total lung dose 

from all fission and activation products was examined for several tests including TEAPOT TURK in 

1955 and LITTLE FELLER I in 1962 (Hicks 1981a,b,c,d). The slope of the trend lines that predict the 

relative importance of 90Sr dose was determined to be 0.0001x for STORAX SMALL BOY, 0.0002x for 

STORAX LITTLE FELLER I (Figure A-3) and TEAPOT TURK (see Figure A-4). Because the slope of 

the trend line is directly proportional to the relative importance of the 90Sr dose to total dose (i.e., the 

larger the slope, the larger the relative importance of 90Sr dose), the tests with the smallest slopes 

result in the highest 

Figure A-2. SMALL BOY – 90Sr fraction of total dose lungs. 

multiplicative correction factors for fission and activation products. Therefore, to ensure the organ 

dose from short-lived fission and activation products is not underestimated, the Hicks (1984) data for 

the STORAX SMALL BOY test were selected to be used to determine the fission and activation 

product dose correction factor. This assumption is justified by the fact that the test was very near the 

last atmospheric test (i.e., STORAX LITTLE FELLER I) and would therefore have been the test most 

likely to produce the short-lived fission and activation product intakes for workers at NTS after 1962 

(the period for which organ dose from environmental intakes is calculated).

Integrating Equation A-2 for SMALL BOY from 0 to 365 days and dividing the result by 365 (the value 

that represents the integrated total dose for 1 year), it was determined that for the first year after 

detonation the lung dose from 90Sr represented 0.0000738 or about 0.00738% of the dose from all 

177 radionuclides. Therefore, the inverse of the value would produce a factor of 13,600 that the lung 

dose from the 90Sr intake (Table A-7) could be multiplied by to account for inhalation dose from shortlived fission and activation products. Similar integrations were performed for subsequent years 

through 1972. 

 

y = 0.0001x + 0.0037

R2

 = 0.9824

0.00

0.05

0.10

0.15

0.20

0.25

0.30

0.35

0.40

0.45

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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Figure A-3. LITTLE FELLER I – 90Sr fraction of total dose to lungs. 

Figure A-4. TURK – 90Sr fraction of total dose to lungs. 

Correction factors to account for inhalation intakes of short-lived fission and activation products have 

been developed for all organs using the Hicks data for STORAX SMALL BOY (Table A-9). These 

correction factors were based on the relationships that Figures A-5 through A-11 show.

 

y = 0.0002x - 0.0124

R2

 = 0.9926

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Totla dose

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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The IMBA computer program was used to determine the organ doses from the scaled intakes in 

Table A-7 for a period of 10 years. These doses were multiplied by the correction factors in Table A-9 

to determine the additional dose that should be added to account for potential dose from inhalation of 

short-lived fission and activation products. The organ-specific fission and activation doses are listed 

in Table A-10. 

Table A-9 Organ-specific inhalation dose fission and activation product correction factors. 

Organ Fission and activation product correction factor

Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Skin Adrenals

Thymus SI 

Spleen Skin 

Muscle Uterus

Pancreas Kidneys

Breast Testes

Esophagus Ovaries

Brain Stomach

Thyroid Gall bladder

730 364 242 182 145 121 104 90.7 80.6 72.5

ULI 458 179 99.2 64.0 45.0 33.4 25.9 20.6 16.8 14.0

Urinary bladder 335 149 91.3 63.6 47.5 37.2 30.1 24.9 21.0 18.0

Lungs 34,900 14,200 7,960 5,150 3,630 2,700 2,100 1,660 1,360 1,130

ET ET1 ET2 LN(TH)

LN(ET)

1,570 827 598 492 438 412 412 412 412 412

LLI 420 142 70.8 42.4 28.2 20.1 15.1 11.7 9.4 7.6

Colon 390 148 79.4 50.0 34.5 25.2 19.3 15.2 12.3 10.2

Liver 9,260 4,620 1,540 1,190 988 858 769 706 661 629

Red bone marrow 37.9 18.2 12.8 10.4 10.4 10.4 10.4 10.4 10.4 10.4

Bone surfaces 78.5 40.1 28.1 22.4 22.4 22.4 22.4 22.4 22.4 22.4

Figure A-5. Strontium-90 fraction of total dose. 

 

y = 4E-06x - 4E-05

R2

 = 0.9936

0.00000

0.00500

0.01000

0.01500

0.02000

0.02500

0.03000

0.03500

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Adrenals Brain Breast Esophagus

Stomach Small Intestine Kidneys Muscle

Ovaries Pancreas Skin Spleen

Testes Thymus Thyroid Uterus

Linear (Kidneys)

ATTACHMENT A

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Figure A-6. Strontium-90 fraction of total dose. 

Figure A-7. Strontium-90 fraction of total dose. 

 

y = 4E-09x2

 + 5E-06x - 0.0002

R2

 = 0.9999

y = 2E-09x2

 + 8E-06x - 0.0002

R2

 = 0.9997

-0.01000

0.00000

0.01000

0.02000

0.03000

0.04000

0.05000

0.06000

0.07000

0.08000

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Urinary Bladder Upper Large Intestine Poly. (Upper Large Intestine) Poly. (Urinary Bladder)

 

y = -2E-11x2 + 2E-07x - 9E-06

R2

 = 0.995

y = -4E-10x2 + 2E-06x - 3E-05

R2

 = 0.9964

0.00000

0.00050

0.00100

0.00150

0.00200

0.00250

0.00300

0.00350

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Extrathoracic Airways Lungs Poly. (Lungs) Poly. (Extrathoracic Airways)

ATTACHMENT A

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Figure A-8. Strontium-90 fraction of total dose. 

Figure A-9. Strontium-90 fraction of total dose to the liver. 

 

y = 9E-09x2

 + 3E-06x + 1E-05

R2

 = 1

y = 6E-09x2

 + 5E-06x - 0.0001

R2

 = 1

0.00000

0.02000

0.04000

0.06000

0.08000

0.10000

0.12000

0.14000

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Lower Large Intestine Colon Poly. (Lower Large Intestine) Poly. (Colon)

 

y = -7E-11x2 + 7E-07x - 3E-05

R2

 = 0.9942

0.000

0.000

0.000

0.001

0.001

0.001

0.001

0.001

0.002

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Totoal Dose

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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Figure A-10. Strontium-90 fraction of total dose to the red marrow. 

Figure A-11. Strontium-90 fraction of total dose to bone surface. 

 

y = -2E-08x2

 + 0.0001x - 0.007

R2

 = 0.9958

0.00

0.05

0.10

0.15

0.20

0.25

0.30

0.35

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

 

y = -6E-09x2

 + 4E-05x - 0.0009

R2

 = 0.9983

-0.01

0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0 500 1000 1500 2000 2500 3000 3500 4000

Days AfterDetonation

Fraction of Total Dose

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

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Table A-10. Inhalation dose from short-lived fission and activation products (rem). Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Bladder 1.86E-04 6.99E-05 3.45E-05 2.26E-05 1.81E-05 1.51E-05 1.29E-05 1.12E-05 9.73E-06 8.53E-06

Brain 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Breast 2.08E-07 3.57E-08 1.65E-08 1.37E-08 1.18E-08 1.01E-08 8.70E-09 7.50E-09 6.49E-09 5.58E-09

Gall bladder 2.08E-07 3.57E-08 1.65E-08 1.37E-08 1.18E-08 1.01E-08 8.70E-09 7.50E-09 6.49E-09 5.58E-09

Heart wall 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Kidney 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Liver 1.93E-03 8.57E-04 3.71E-04 2.70E-04 2.33E-04 2.07E-04 1.85E-04 1.67E-04 1.52E-04 1.39E-04

Muscle 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Ovaries 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Pancreas 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Testes 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Thyroid 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

RBM 4.08E-04 7.00E-04 6.65E-04 6.09E-04 5.59E-04 5.18E-04 4.84E-04 4.57E-04 4.35E-04 4.16E-04

Bone surface 1.23E-03 2.03E-03 2.00E-03 1.91E-03 1.83E-03 1.77E-03 1.72E-03 1.69E-03 1.66E-03 1.64E-03

Stomach 1.69E-04 7.22E-05 4.14E-05 3.02E-05 2.57E-05 2.26E-05 2.01E-05 1.81E-05 1.63E-05 1.47E-05

SI 1.85E-04 7.67E-05 4.36E-05 3.15E-05 2.68E-05 2.35E-05 2.09E-05 1.87E-05 1.69E-05 1.53E-05

ULI 4.70E-04 1.36E-04 5.98E-05 3.59E-05 2.73E-05 2.19E-05 1.81E-05 1.52E-05 1.29E-05 1.11E-05

LLI 1.38E-03 3.26E-04 1.24E-04 6.66E-05 4.74E-05 3.63E-05 2.90E-05 2.37E-05 1.98E-05 1.67E-05

Skin 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Spleen 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Thymus 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Uterus 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

ET 2.22E-03 6.97E-04 3.78E-04 2.60E-04 2.32E-04 2.19E-04 2.17E-04 2.15E-04 2.14E-04 2.13E-04

Lung 7.67E-03 2.96E-03 1.55E-03 1.08E-03 8.84E-04 7.46E-04 6.40E-04 5.52E-04 4.80E-04 4.18E-04

Colon 7.82E-04 2.08E-04 8.57E-05 4.85E-05 3.57E-05 2.78E-05 2.26E-05 1.86E-05 1.57E-05 1.34E-05

ET1 1.63E+00 4.74E-01 2.48E-01 1.64E-01 1.46E-01 1.38E-01 1.38E-01 1.38E-01 1.38E-01 1.38E-01

ET2 5.83E-04 2.24E-04 1.30E-04 9.60E-05 8.63E-05 8.10E-05 7.92E-05 7.78E-05 7.66E-05 7.55E-05

LN(ET) 3.27E-04 1.50E-04 9.15E-05 7.03E-05 6.34E-05 5.95E-05 5.77E-05 5.63E-05 5.51E-05 5.40E-05

LN(TH) 3.27E-04 1.50E-04 9.15E-05 7.03E-05 6.34E-05 5.95E-05 5.77E-05 5.63E-05 5.51E-05 5.40E-05

Esophagus 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

The application of the early fission and activation product correction factor through 1972 intakes is 

favorable to claimants because the Hicks (1984) data verifies that after 10 years, more than 90% of 

the total dose for all organs is delivered by 241Am, 137Cs, and/or 90Sr. Further, as times after 

detonation become greater, the relative importance of 241Am, 137Cs, and 90Sr becomes greater and, 

because the organ doses from these radionuclides are already accounted for by the scaled intakes 

(Table A-7), the dose from these radionuclides is doubled. 

A.7 Ingestion Pathway

To account for potential intakes from inadvertent ingestion of contaminated soil, the area-specific 

radionuclide soil deposition data in Table A-5 were converted to volumetric data (i.e., Bq/mg) by 

assuming a radionuclide relaxation depth of 2.3 cm and a soil density of 1.5 g/cm3 (DOE 2003). The 

area-specific radionuclide soil concentrations are listed in Table A-11.

Table A-11. Radionuclide soil concentration by area (Bq/g). 

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 0.069 0.127 0.375 0.682 0.263 0.456 1.005 0.014 0.391

2 0.064 0.226 0.463 1.001 0.963 1.882 1.262 0.420

3 0.062 0.050 0.475 0.509 0.294 0.824 0.989 0.012 0.321

4 0.179 0.420 1.036 1.644 0.593 0.655 1.010 0.259

5 0.090 0.018 0.686 3.401 0.109 0.250 6.122 0.259

6 0.023 0.053 0.108 0.102 0.069 0.087 0.000 0.00

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

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Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

7 0.049 0.016 0.344 0.852 0.213 0.384 2.024 0.039 0.322

8 0.530 0.297 3.280 6.741 2.389 1.450 0.562 0.894

9 0.091 0.057 1.844 0.575 0.344 0.524 2.042 0.038 0.311

10 0.411 0.491 2.279 7.973 3.321 2.217 0.195 0.056 5.177

11 0.357 0.065 3.005 0.099 0.060

12 0.062 0.111 0.408 0.498 0.399 0.346

15 0.098 0.114 0.740 0.140 0.426 0.502

16 0.021 0.054 0.107 0.115 0.160 0.209

17 0.039 0.074 0.238 0.524 0.378 0.488

18 0.301 0.106 1.518 0.421 0.290 0.502 0.072 0.014 0.607

19 0.061 0.111 0.391 0.122 0.192 0.169

20 1.607 2.500 2.741 20.946 0.701 0.559 3.723 0.969 16.031

25 0.176 0.090 0.789

26

30 4.620 7.749 19.340 43.835 3.953 3.493 4.143 1.252 13.804

If the assumption is made that the workers ingested 100 mg of soil each day, which is favorable to 

claimants [EPA (1989) recommends a value of 50 mg/d], and that full-time employment was 250 d/yr, 

annual ingestion can be calculated. The area-specific annual ingestion rates are listed in Table A-12. 

Table A-12. Area-specific and maximum annual ingestion rates (Bq/yr). 

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 1.72 3.17 9.38 17.06 6.56 11.41 25.12 0.35 9.77

2 1.59 5.64 11.57 25.03 24.08 47.06 31.54 10.51

3 1.54 1.24 11.87 12.72 7.34 20.59 24.73 0.29 8.01

4 4.47 10.49 25.90 41.10 14.83 16.37 25.24 6.47

5 2.24 0.45 17.15 85.03 2.73 6.25 153.05 6.48 0.00

6 0.57 1.32 2.69 2.54 1.71 2.18

7 1.23 0.40 8.59 21.29 5.33 9.61 50.59 0.97 8.05

8 13.24 7.43 81.99 168.52 59.73 36.25 14.05 0.00 22.34

9 2.27 1.42 46.11 14.38 8.60 13.10 51.04 0.94 7.76

10 10.29 12.27 56.98 199.31 83.02 55.42 4.88 1.41 129.41

11 8.93 1.61 75.12 0.00 2.47 1.51

12 1.56 2.77 10.20 12.45 9.98 8.65

15 2.45 2.85 18.49 3.49 10.64 12.56

16 0.53 1.35 2.68 2.87 4.01 5.21

17 0.97 1.85 5.94 13.09 9.44 12.19

18 7.54 2.65 37.95 10.54 7.24 12.55 1.79 0.34 15.17

19 1.53 2.79 9.78 3.05 4.80 4.21

20 40.17 62.49 68.51 523.64 17.54 13.98 93.06 24.23 400.77

25 4.39 2.24 19.73

26

30 115.50 193.72 483.50 1095.89 98.83 87.33 103.56 31.30 345.10

Maxa 40.17 62.49 81.99 523.64 83.02 55.42 153.05 24.23 400.77

a. Maximum value with Area 30 excluded.

For most radionuclides, Area 30 provided the highest areal deposition and resultant intakes. Area 30 

is relatively small (150 km2

) and inaccessible, and is on the Western edge of NTS. It has rugged 

terrain and includes the northern reaches of Fortymile Canyon. In 1968, it was the site of Project 

ATTACHMENT A

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Page 20 of 27
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BUGGY, the first nuclear row-charge experiment in the PLOWSHARE Program. As a result of the 

test, a trench 255 m long, 77 m wide, and 206 m deep was created. The test resulted in large 

quantities of vitrified sand. Because of the bias that is introduced when Area 30 is included, the 

maximum annual intakes in Table A-12 have been provided with Area 30 areal concentrations 

excluded. The Area 30 intakes should be used only if it can be determined that the worker was 

assigned to and primarily worked in Area 30.

To determine the relative importance of the ingestion pathway, a hypothetical ingestion scenario was 

used that assumed 30 years of the Table A-12 maximum intakes and used the IMBA computer 

program to determine organ doses from these intakes. The annual organ doses from ingestion of 

238,239Pu and 241Am greater than 0.001 rem are listed in Table A-13. With the exception of the red 

bone marrow and bone surfaces, the annual doses from the Table A-12 maximum ingestion rates of 

60Co, 137Cs, and 152,154,155Eu were all less than 0.001 rem. For the red bone marrow and bone 

surfaces, the annual doses were all less than 0.002 rem.

Table A-13. Annual organ doses from ingestion of 238,239Pu and 241Am 

(rem). 

Year Liver Ovaries Testes RBM

Bone 

surface LLI

1963 0.002 0.001

1964 0.001 0.007 0.001

1965 0.002 0.001 0.011 0.001

1966 0.003 0.001 0.016 0.001

1967 0.004 0.002 0.020 0.001

1968 0.005 0.002 0.024 0.001

1969 0.005 0.002 0.028 0.001

1970 0.006 0.002 0.032 0.001

1971 0.007 0.003 0.035 0.001

1972 0.008 0.003 0.039 0.001

1973 0.008 0.003 0.043 0.001

1974 0.009 0.003 0.046 0.001

1975 0.010 0.003 0.050 0.001

1976 0.010 0.004 0.053 0.001

1977 0.011 0.004 0.057 0.001

1978 0.011 0.004 0.060 0.001

1979 0.012 0.004 0.063 0.001

1980 0.013 0.001 0.001 0.004 0.066 0.001

1981 0.013 0.001 0.001 0.004 0.070 0.001

1982 0.014 0.001 0.001 0.004 0.073 0.001

1983 0.014 0.001 0.001 0.004 0.076 0.001

1984 0.015 0.001 0.001 0.005 0.079 0.001

1985 0.015 0.001 0.001 0.005 0.082 0.001

1986 0.016 0.001 0.001 0.005 0.085 0.001

1987 0.016 0.001 0.001 0.005 0.088 0.001

1988 0.017 0.001 0.001 0.005 0.091 0.001

1989 0.017 0.001 0.001 0.005 0.094 0.001

1990 0.018 0.001 0.001 0.005 0.096 0.001

1991 0.018 0.002 0.002 0.005 0.099 0.001

1992 0.019 0.002 0.002 0.005 0.102 0.001

1993 0.019 0.002 0.002 0.005 0.105 0.001

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

Page 21 of 27
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 79 of 88

Year Liver Ovaries Testes RBM

Bone 

surface LLI

1994 0.019 0.002 0.002 0.005 0.105

1995 0.019 0.002 0.002 0.005 0.103

1996 0.018 0.002 0.002 0.005 0.101

1997 0.018 0.002 0.002 0.004 0.099

1998 0.017 0.002 0.002 0.004 0.098

1999 0.017 0.001 0.001 0.004 0.096

2000 0.017 0.001 0.001 0.004 0.095

2001 0.016 0.001 0.001 0.004 0.093

2002 0.016 0.001 0.001 0.003 0.092

2003 0.016 0.001 0.001 0.003 0.090

2004 0.015 0.001 0.001 0.003 0.089

2005 0.015 0.001 0.001 0.003 0.088

2006 0.015 0.001 0.001 0.003 0.086

2007 0.014 0.001 0.001 0.003 0.085

As with the inhalation intakes discussed in Section A.5, ingestion doses need to be adjusted for 

potential dose from short-lived fission and activation products that are no longer persistent in NTS 

soils in measurable amounts. The organ-specific fission and activation product correction factors 

were developed (for the reasons discussed in Section A.5) based on the relative contribution of 90Sr to 

the total ingestion dose using the Hicks (1981c) data from the STORAX SMALL BOY test in July 

1962. As in Section A.5 for inhalation dose, organ-specific relationships were developed for the 90Sr 

fraction of total dose. These relationships are shown in Figures A-12 through A-17. With the 

exception of fractional dose to bone surfaces and red bone marrow, these relationships were valid out 

to 10 years after detonation. For bone surfaces and red bone marrow, the time-dependant 

relationships were determined for the first year and for the last 9 years. 

To account for fractionation, the refractory elements included in the STORAX SMALL BOY Hicks 

(1981c) data were multiplied by a factor of 2. Increasing the abundance of the refractory elements 

provides a more reasonable estimate of their relative contribution to total dose. As for the inhalation 

correction factors, these radionuclide- and time-dependant relative abundances are multiplied by their 

organ-specific ingestion dose conversion factors (Bunker 1999) to determine the radionuclide-specific, 

relative importance to total dose as a function of time after detonation. Because of the large 

difference between the 50-year committed dose conversion factor and the annual dose conversion 

factor for 241Am, the IMBA computer program was used to calculate the annual ingestion dose 

conversion factors, which were then used rather than the 50-year committed dose conversion factors. 

The same method was used for 90Sr annual dose conversion factors. To ensure that organ doses 

were not underestimated, the f1 factor for each radionuclide was chosen to provide the largest dose to 

the specific organ of interest.

The ingestion correction factors were evaluated over a 10-year period. This is because 144Ce and 

106Ru continue to provide relatively large ingestion doses through 5 years after detonation. Therefore, 

to capture their contributions to total dose, the correction factor integration times were extended to 10 

years if the relative contribution of 144Ce and 106Ru to total dose was less than 3% for all organs. In 

addition, at the end of 10 years, more than 95% of the total ingestion dose is delivered by 90Sr and 

137Cs, their dose accounted for by the ingestion intakes in Table A-12. 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

Page 22 of 27
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 80 of 88

Figure A-12. Strontium-90 fraction of total ingestion dose. 

Figure A-13. Strontium-90 fraction of total ingestion dose to red bone marrow. 

 

y = 2E-05x - 0.001

R2

 = 0.9845

y = 0.0684Ln(x) - 0.1775

R2

 = 0.88

0.00000

0.10000

0.20000

0.30000

0.40000

0.50000

0.60000

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Bone Surface Upper Large Intestine Lower Large Intestine

Colon Red Marrow Linear (Upper Large Intestine)

Log. (Bone Surface)

 

y = 0.0008x - 0.0107

R2

 = 0.9879

-0.05000

0.00000

0.05000

0.10000

0.15000

0.20000

0.25000

0.30000

0 50 100 150 200 250 300 350 400

Days After Detonation

Fraction of Total Dose

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

Page 23 of 27
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 81 of 88

Figure A-14. Strontium-90 fraction of total ingestion dose to red bone marrow. 

Figure A-15. Strontium-90 fraction of total ingestion dose. 

 

y = -4E-08x2

 + 0.0002x + 0.196

R2

 = 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

Page 24 of 27
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 82 of 88

Figure A-16. Strontium-90 fraction of ingestion dose to the upper large intestine. 

Figure A-17. Strontium-90 fraction of total ingestion dose. 

 

 

y = -2E-09x2 + 3E-05x - 0.0022

R2

 = 0.9908

0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

 

y = 4E-05x - 0.0024

R2

 = 0.9929

0.00

0.02

0.04

0.06

0.08

0.10

0.12

0.14

0.16

0.18

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Lower Large Intestine Colon Linear (Colon)

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

Page 25 of 27
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 83 of 88

Using the relationships shown in Figures A-12 through A-17, short-lived fission and activation 

ingestion correction factors were developed (Table A-14). As with the inhalation fission and activation 

factors, these correction factors are multiplied by the organ-specific 90Sr annual doses (from the 

intakes in Table A-12 [i.e., 55.42 Bq/yr]) to calculate additional ingestion dose from short-lived fission 

and products. 

Table A-14. Organ-specific ingestion fission and activation correction factors. 

Organ Correction factor

Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals, breast 

Brain, skin, bladder

Stomach, kidneys, muscle, 

pancreas, brain, 

esophagus, SI, liver, 

ovaries ET ET1

ET2 LM(ET) LN(TH), lungs, 

skin, spleen, testes, 

thymus, thyroid, uterus

Gall bladder

416 219 155 123 106 95.0 88.1 84.0 82.0 81.8

ULI 514 184 95.2 58.2 39.3 18.6 16.3 14.6 13.4 12.4

Bone Surface/RBM 3.8 3.1 2.7 2.5 2.3 2.3 2.3 2.3 2.3 2.3

LLI/Colon 417 208 138 25.6 20.4 17.4 14.6 12.8 11.3 10.2

To simplify the application of organ-specific ingestion dose from short-lived fission and activation 

products, the IMBA computer program was used to determine the organ-specific annual doses for the 

90Sr intake of 55.42 Bq/yr for 1963 through 1972 (Table A-15). These doses are multiplied by the 

Table A-14 organ-specific correction factors to provide the annual doses from ingestion of short-lived 

fission and activation products (Table A-16).

Table A-15. Organ-specific annual ingestion doses (rem) for the 90Sr intake of 55.42 Bq/yr. Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Urinary 

bladder 5.68E-06 6.42E-06 6.67E-06 6.87E-06 7.04E-06 7.18E-06 7.31E-06 7.41E-06 7.50E-06 7.58E-06

Bone surface 1.58E-04 3.74E-04 5.47E-04 6.98E-04 8.32E-04 9.50E-04 1.06E-03 1.15E-03 1.23E-03 1.31E-03

Brain 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Breast 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Esophagus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Stomach 3.42E-06 3.78E-06 3.95E-06 4.09E-06 4.21E-06 4.31E-06 4.40E-06 4.47E-06 4.54E-06 4.59E-06

Small 

intestine 4.71E-06 5.07E-06 5.24E-06 5.38E-06 5.50E-06 5.60E-06 5.69E-06 5.76E-06 5.83E-06 5.88E-06

Upper large 

intestine 3.42E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05

Lower large 

intestine 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04

Colon 4.36E-03 2.26E-03 1.51E-03 1.14E-03 9.16E-04 7.71E-04 6.61E-04 5.84E-04 5.21E-04 4.71E-04

Kidneys 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Liver 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Muscle 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Ovaries 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Pancreas 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Red marrow 1.08E-04 2.66E-04 3.87E-04 4.86E-04 5.69E-04 6.37E-04 6.94E-04 7.42E-04 7.81E-04 8.14E-04

Extrathoracic 

airways 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Lungs 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

Page 26 of 27
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 84 of 88

Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Skin 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Spleen 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Testes 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Thymus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Thyroid 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Uterus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Table A-16. Organ-specific doses (rem) from ingestion of fission and activation products. Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Bladder 2.36E-03 1.55E-03 1.15E-03 9.50E-04 8.45E-04 7.76E-04 7.31E-04 7.02E-04 6.85E-04 2.21E-04

Brain 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Breast 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Gall bladder 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Heart wall 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Kidney 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Liver 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Muscle 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Ovaries 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Pancreas 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Testes 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Thyroid 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

RBM 4.12E-04 2.66E-04 3.87E-04 4.86E-04 5.69E-04 6.37E-04 6.94E-04 7.42E-04 7.81E-04 8.14E-04

Bone surface 6.01E-04 1.31E-03 1.75E-03 2.09E-03 2.35E-03 2.56E-03 2.76E-03 2.94E-03 3.11E-03 3.05E-03

Stomach 1.42E-03 9.00E-04 6.79E-04 5.71E-04 5.13E-04 4.74E-04 4.48E-04 4.31E-04 4.21E-04 1.41E-04

SI 1.96E-03 1.18E-03 8.79E-04 7.29E-04 6.49E-04 5.96E-04 5.61E-04 5.39E-04 5.27E-04 1.43E-04

ULI 1.53E-02 5.87E-03 3.11E-03 1.96E-03 1.36E-03 7.26E-04 6.30E-04 5.64E-04 5.14E-04 1.34E-04

LLI 4.84E-02 2.50E-02 1.66E-02 3.47E-03 2.66E-03 2.23E-03 1.92E-03 1.68E-03 1.49E-03 1.34E-03

Skin 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Spleen 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Thymus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Uterus 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

ET 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Lung 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Colon 2.80E-02 1.45E-02 9.66E-03 2.07E-03 1.59E-03 1.33E-03 1.15E-03 1.01E-03 8.94E-04 8.06E-04

ET1 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

ET2 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

LN(ET) 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

LN(TH) 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Esophagus 2.89E-04 2.07E-04 1.62E-04 1.41E-04 1.29E-04 1.21E-04 1.16E-04 1.13E-04 1.11E-04 5.02E-05

A.8 Instruction to Dose Reconstructors for Assignment of Environmental Intakes

With the exception of cases that can be worked with the use of the bounding assumptions from 

ORAUT-OTIB-0018 (ORAUT 2005), environmental inhalation and ingestion intakes in Tables A-7 and 

A-12, respectively, should be applied starting in 1963. In addition, for applicable years of employment 

and affected organs, the annual doses in Tables A-10 and A-16 should be applied to account for dose 

from inhalation and ingestion of short-lived fission and activation products. These intakes and 

resultant doses should be entered into the IREP analysis with a constant distribution because they are 

reasonable overestimates of the actual intakes and doses. It should be noted that doses that are less 

than 0.001 rem do not need to be entered into the IREP analysis because they would not affect the 

overall dose to the affected organ. The organs that have dose equal to or greater than 0.001 rem and 

the resultant dose are given in Tables 4-9 and 4-14. Only these doses need to be entered into the 

IREP analysis and they should be entered as 30-to-250-keV photons with a constant distribution. 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA 

Page 27 of 27
Document No. ORAUT-TKBS-0008-4 Revision No. 03 Effective Date: 08/24/2012 Page 85 of 88

ATTACHMENT B

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

Page 1 of 4

In the previous sections, methods were presented to account for organ dose from ambient 

environmental inhalation and ingestion intakes of radioactive materials in the NTS atmosphere and 

soils. To illustrate the importance of these pathways, the IMBA computer program was used to 

determine organ doses from 30 years of the inhalation intakes in Table A-7 and the ingestion intakes 

in Table A-12. These 30-year organ doses and the fractional contribution for each of the 

radionuclides persistent in NTS soils are listed in Table B-1. To correct for exposure to short-lived 

fission and activation, the dose reconstructor should add annual doses that were greater than 

0.001 rem from Tables 4-9 and 4-14 as 30-to-250-keV photons with a constant distribution. The 

doses shown in table B-1 are for purposes of illustration only and should not be entered into the IREP 

analysis. These doses will be accounted for by the assignment of the intakes shown in Tables 4-7 

and 4-11 (MAX) and repeated in Tables A-7 and A-12 (MAX) for all years of employment.
Document No. ORAUT-TKBS-0008-4 Revision No. 0

3 Effective Date: 08/2

4/2012 Page 86 of 88

ATTACHMENT B

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

Page 2 of 4

Table B-1. Total annual organ doses greater than 0.0001 rem from 30 years of inhalation and ingestion intakes (rem). Year Radiation Type Liver Ovaries Testes RBM Bone surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

1963 Alpha 0.001 0.006 0.001 0.004 0.007 0.001 0.005 0.004 0.001

1964 Alpha 0.003 0.002 0.017 0.001 0.009 0.006 0.001 0.003 0.009 0.001 0.003

1965 Alpha 0.005 0.002 0.026 0.001 0.009 0.005 0.001 0.002 0.009 0.001 0.004

1966 Alpha 0.006 0.003 0.033 0.001 0.009 0.004 0.001 0.002 0.009 0.002 0.006

1967 Alpha 0.008 0.003 0.040 0.001 0.009 0.004 0.001 0.002 0.009 0.003 0.007

1968 Alpha 0.009 0.004 0.046 0.001 0.008 0.004 0.001 0.002 0.008 0.003 0.008

1969 Alpha 0.010 0.004 0.052 0.001 0.008 0.004 0.001 0.002 0.008 0.004 0.009

1970 Alpha 0.011 0.004 0.058 0.001 0.008 0.004 0.001 0.002 0.008 0.004 0.011

1971 Alpha 0.012 0.005 0.065 0.001 0.008 0.004 0.001 0.002 0.008 0.005 0.012

1972 Alpha 0.013 0.001 0.001 0.005 0.070 0.001 0.008 0.004 0.001 0.002 0.008 0.005 0.013

1973 Alpha 0.015 0.001 0.001 0.005 0.076 0.001 0.008 0.004 0.001 0.002 0.008 0.006 0.014

1974 Alpha 0.016 0.001 0.001 0.006 0.082 0.001 0.008 0.004 0.001 0.002 0.008 0.006 0.016

1975 Alpha 0.017 0.001 0.001 0.006 0.088 0.001 0.008 0.005 0.001 0.002 0.008 0.007 0.017

1976 Alpha 0.018 0.001 0.001 0.006 0.093 0.001 0.008 0.005 0.001 0.002 0.008 0.007 0.018

1977 Alpha 0.019 0.001 0.001 0.006 0.098 0.001 0.008 0.005 0.001 0.002 0.008 0.007 0.019

1978 Alpha 0.020 0.002 0.002 0.006 0.104 0.001 0.008 0.005 0.001 0.002 0.008 0.008 0.020

1979 Alpha 0.021 0.002 0.002 0.007 0.109 0.001 0.008 0.005 0.001 0.002 0.008 0.008 0.022

1980 Alpha 0.021 0.002 0.002 0.007 0.114 0.001 0.008 0.005 0.001 0.002 0.008 0.008 0.023

1981 Alpha 0.022 0.002 0.002 0.007 0.119 0.001 0.008 0.005 0.001 0.002 0.008 0.009 0.024

1982 Alpha 0.023 0.002 0.002 0.007 0.124 0.001 0.008 0.005 0.001 0.002 0.008 0.009 0.025

1983 Alpha 0.024 0.002 0.002 0.007 0.129 0.001 0.008 0.005 0.001 0.002 0.008 0.009 0.026

1984 Alpha 0.025 0.002 0.002 0.008 0.134 0.001 0.008 0.005 0.001 0.002 0.008 0.009 0.027

1985 Alpha 0.026 0.002 0.002 0.008 0.139 0.001 0.008 0.005 0.001 0.002 0.008 0.010 0.028

1986 Alpha 0.027 0.002 0.002 0.008 0.144 0.001 0.008 0.005 0.001 0.002 0.008 0.010 0.029

1987 Alpha 0.027 0.002 0.002 0.008 0.149 0.001 0.008 0.005 0.001 0.002 0.008 0.010 0.030

1988 Alpha 0.028 0.002 0.002 0.008 0.153 0.001 0.008 0.005 0.001 0.002 0.008 0.010 0.031

1989 Alpha 0.029 0.002 0.002 0.008 0.158 0.001 0.008 0.005 0.001 0.002 0.008 0.011 0.032

1990 Alpha 0.030 0.002 0.002 0.008 0.162 0.001 0.008 0.005 0.001 0.002 0.008 0.011 0.033

1991 Alpha 0.030 0.003 0.003 0.009 0.167 0.001 0.008 0.005 0.001 0.002 0.008 0.011 0.034

1992 Alpha 0.031 0.003 0.003 0.009 0.171 0.001 0.008 0.005 0.001 0.002 0.008 0.011 0.035

1993 Alpha 0.031 0.003 0.003 0.009 0.172 0.006 0.003 0.006 0.011 0.035

1994 Alpha 0.030 0.003 0.003 0.008 0.169 0.004 0.002 0.004 0.011 0.035

1995 Alpha 0.030 0.003 0.003 0.008 0.166 0.003 0.002 0.003 0.011 0.035

1996 Alpha 0.029 0.003 0.003 0.007 0.163 0.002 0.001 0.002 0.011 0.035

1997 Alpha 0.028 0.002 0.003 0.007 0.160 0.001 0.001 0.001 0.011 0.035

1998 Alpha 0.028 0.002 0.002 0.006 0.158 0.010 0.034

1999 Alpha 0.027 0.002 0.002 0.006 0.155 0.010 0.034

2000 Alpha 0.026 0.002 0.002 0.006 0.153 0.010 0.033
Document No. ORAUT-TKBS-0008-4 Revision No. 0

3 Effective Date: 08/2

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ATTACHMENT B

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

Page 3 of 4

Year Radiation Type Liver Ovaries Testes RBM Bone surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

2001 Alpha 0.026 0.002 0.002 0.006 0.151 0.009 0.033

2002 Alpha 0.025 0.002 0.002 0.005 0.148 0.009 0.032

2003 Alpha 0.025 0.002 0.002 0.005 0.146 0.009 0.032

2004 Alpha 0.024 0.002 0.002 0.005 0.144 0.008 0.031

2005 Alpha 0.024 0.002 0.002 0.005 0.142 0.008 0.031

2006 Alpha 0.023 0.002 0.002 0.005 0.140 0.008 0.030

2007 Alpha 0.023 0.002 0.002 0.005 0.138 0.007 0.029

2008 Alpha 0.022 0.002 0.002 0.004 0.136 0.007 0.029

2009 Alpha 0.022 0.002 0.002 0.004 0.134 0.007 0.028

2010 Alpha 0.021 0.002 0.002 0.004 0.132 0.007 0.027

2011 Alpha 0.021 0.002 0.002 0.004 0.130 0.006 0.026

2012 Alpha 0.021 0.002 0.002 0.004 0.128 0.006 0.026

1963 Photons E>250 keV 0.002

1964 Photons E>250 keV 0.001

1965 Photons E>250 keV 0.001

1963 Electrons E>15 keV 0.002

1964 Electrons E>15 keV 0.001

1965 Electrons E>15 keV

1966 Electrons E>15 keV

1967 Electrons E>15 keV 0.001

1968 Electrons E>15 keV 0.001

1969 Electrons E>15 keV 0.001

1970 Electrons E>15 keV 0.001

1971 Electrons E>15 keV 0.001

1972 Electrons E>15 keV 0.002

1973 Electrons E>15 keV 0.002

1974 Electrons E>15 keV 0.001 0.002

1975 Electrons E>15 keV 0.001 0.002

1976 Electrons E>15 keV 0.001 0.002

1977 Electrons E>15 keV 0.001 0.002

1978 Electrons E>15 keV 0.001 0.002

1979 Electrons E>15 keV 0.001 0.002

1980 Electrons E>15 keV 0.001 0.002

1981 Electrons E>15 keV 0.001 0.002

1982 Electrons E>15 keV 0.001 0.002

1983 Electrons E>15 keV 0.001 0.002

1984 Electrons E>15 keV 0.001 0.002

1985 Electrons E>15 keV 0.001 0.002

1986 Electrons E>15 keV 0.001 0.002
Document No. ORAUT-TKBS-0008-4 Revision No. 0

3 Effective Date: 08/2

4/2012 Page 88 of 88

ATTACHMENT B

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

Page 4 of 4

Year Radiation Type Liver Ovaries Testes RBM Bone surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

1987 Electrons E>15 keV 0.001 0.002

1988 Electrons E>15 keV 0.001 0.002

1989 Electrons E>15 keV 0.001 0.002

1990 Electrons E>15 keV 0.001 0.002

1991 Electrons E>15 keV 0.001 0.002

1992 Electrons E>15 keV 0.001 0.002

1993 Electrons E>15 keV 0.002

1994 Electrons E>15 keV 0.002

1995 Electrons E>15 keV 0.002

1996 Electrons E>15 keV 0.001

1997 Electrons E>15 keV 0.001

1998 Electrons E>15 keV 0.001

1999 Electrons E>15 keV 0.001

2000 Electrons E>15 keV 0.001